• Korean Chemical Engineering Research
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• v.50 no.4
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• pp.696-701
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• 2012

Non-aqueous processes have been developed for stable management and reuse of spent fuels. Nowadays, a plan for the management of spent fuel is being sought focusing on a non-aqueous process in Korea. Named as pyroprocessing, it includes an electrolytic reduction process using molten salt at high temperature for the spent fuels, which provides metallic product for a following electro-refining process. The electrolytic reduction process utilizes electrochemical reaction producing Li to convert oxides into metals in high temperature LiCl medium. Various kinds of elements in the spent fuels would be distributed in the system according to their respective reactivity with the reductant, Li, and the medium, LiCl. This study elucidates the reactions of the elements to understand the behavior of composite elements on the spent fuels by thermodynamic calculations. Uranium and transuranic are reduced into their metallic forms while rare-earth oxides, except for Eu, are stable against the reaction at a process temperature. This study also covers the tendency of reactions with respect to the temperature and, finally, estimates radioactivity and heat load on the distributed phases based on the reference spent fuel characteristics.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.383-395
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• 2023

Since 1997, the Republic of Korea (ROK) has been developing pyro-processing (Pyro) technology to reduce the disposal burden of high-level radioactive waste by recycling spent nuclear fuel (SNF). Compared to plutonium and uranium extraction process, Korean Pyro technology has relatively excellent proliferation resistance that cannot separate pure plutonium owing to its intrinsic characteristics. Regarding Pyro technology development of ROK, the Bush administration considered that Pyro is not reprocessing under the Global Nuclear Energy Partnership, whereas the Obama administration considered that Pyro is subject to reprocessing. However, the Bush and Obama administrations did not allow ROK to conduct full Pyro activities using SNF, even though ROK had faithfully complied with international nonproliferation obligations. This is because the US nuclear nonproliferation policy to prevent the spread of sensitive technologies, such as enrichment and reprocessing, has a strong effect on ROK, unlike Japan, on a bilateral level beyond the NPT regime for non-proliferation of nuclear weapons.

• Journal of Institute of Convergence Technology
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• v.4 no.2
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• pp.47-50
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• 2014

Voloxidation is a process for converting $UO_2$ into $U_3O_8$ while removing some volatile products in spent fuels (SF). Various oxidative gas conditions including air and mixture of Ar and $O_2$ could be adopted for the process. The gas flows into a reactor under high temperature ($>500^{\circ}C$) and components of SF are reacted with the gas. SF is composed of various components such as actinides, lanthanides, and alkali metals. Therefore, it is of significance to understand their behavior during the reactions for process development. However, due to the limit of available experiments, phase diagram analysis should be preceded. TPP diagram is constructed with respect to temperature-pressure-pressure. It shows a stable phase depending on partial pressures of gas components as well as temperature. In this work, we investigated TPP diagrams for actinides, lanthanides and other oxides to determine stable oxide forms under different gas conditions. The results would be used to set up a material balance under a pyroprocessing scheme of SF and compare the gas conditions for the optimization of fission products removal.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.19-32
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• 2010

Electrolytic reduction technology is essential for the purpose of adopting pyroprocessing into spent oxide fuel as an alternative option in a back-end fuel cycle. Spent fuel consists of various metal oxides, and each metal oxide releases an oxygen element depending on its chemical characteristic during the electrolytic reduction process. In the present work, an electrolytic reduction behavior was estimated for voloxidized spent fuel based on the assumption that each metal-oxygen system is independent and behaves as an ideal solid solution. The electrolytic reduction was considered as a combination of a Li recovery and chemical reactions between the metal oxides such as uranium oxide and the produced Li metal. The calculated result revealed that most of the metal oxides were reduced by the process. It was evaluated that a reduced fraction of lanthanide oxides increased with a decreasing $Li_2O$ concentration. However, most of the lanthanides were expected to be stable in their oxide forms. In addition, a semi-empirical model for describing $U_3O_8$ electrolytic reduction behavior was proposed by considering Li diffusion and a chemical reaction between $U_3O_8$ and Li. Experimental data was used to determine model parameters and, then, the model was applied to calculate the reduction yield with time and to estimate the required time for a 99.9% reduction.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.148-164
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• 2017

The focus on the issues surrounding spent nuclear fuel and lifetime extension of old nuclear power plants continues to grow nowadays. A transparent decision-making process to identify the best suitable nuclear fuel cycle (NFC) is considered to be the key task in the current situation. Through this study, an attempt is made to develop an equilibrium model for the NFC to calculate the material flows based on 1 TWh of electricity production, and to perform integrated multicriteria decision-making method analyses via the analytic hierarchy process technique for order of preference by similarity to ideal solution, preference ranking organization method for enrichment evaluation, and multiattribute utility theory methods. This comparative study is aimed at screening and ranking the three selected NFC options against five aspects: sustainability, environmental friendliness, economics, proliferation resistance, and technical feasibility. The selected fuel cycle options include pressurized water reactor (PWR) once-through cycle, PWR mixed oxide cycle, or pyroprocessing sodium-cooled fast reactor cycle. A sensitivity analysis was performed to prove the robustness of the results and explore the influence of criteria on the obtained ranking. As a result of the comparative analysis, the pyroprocessing sodium-cooled fast reactor cycle is determined to be the most competitive option among the NFC scenarios.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.1
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• pp.83-90
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• 2017

The pyroprocessing of spent nuclear fuel generates LiCl-KCl eutectic waste salt containing radioactive rare earth nuclides. It is necessary to develop a simple process for the treatment of LiCl-KCl eutectic waste in a hot-cell facility. In this study, capture and solidification of a rare earth nuclide (Nd) in LiCl-KCl eutectic salt using an inorganic composite with a $Li_2O-Al_2O_3-SiO_2-B_2O_3$ system was conducted to simplify the existing separation and solidification process of rare earth nuclides in LiCl-KCl eutectic waste salt from the pyroprocessing of spent nuclear fuel. More than 98wt% of Nd in LiCl-KCl eutectic salt was captured when the mass ratio of the composite was 0.67 over $NdCl_3$ in the eutectic salt. The content of $Nd_2O_3$ in the Nd captured-composite reached about 50wt%, and this composite was directly fabricated into a homogeneous and chemical resistant glass waste in a monolithic form. These results will be utilized in designing a process to simplify the existing separation and solidification process.

• 제어로봇시스템학회:학술대회논문집
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• 2005.06a
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• pp.2226-2230
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• 2005

The Advanced Spent Fuel Conditioning Process (ACP), which is a pre-disposal treatment process for spent fuel is being developed at the Korea Atomic Energy Research Institute (KAERI). The ACP equipment is operated in an intense radiation field as well as in a high temperature. Thus, the equipment is designed in consideration of the remote handling and maintenance. This paper describes a Bridge Transported Bilateral Force-Reflecting Servo-Manipulator (BTSM) system, which is being developed to overcome the limitation of access that is a drawback of the mechanical Master-Slave Manipulators (MSMs), which are mounted on the ACP hot cell wall for the operation and the maintenance of the ACP equipment. The BTSM system was manufactured and temporally installed at the mockup to test its performance. The manufactured BTSM system will be installed at the ACP hot cell on June 2005 after the accomplishment of the performance test. The BTSM system consists of four components: a transporter with a telescoping tubeset, a slave manipulator, a master manipulator, and a remote control system. This system will highly increase the volume of coverage for the operation and maintenance of the ACP equipment.

• Journal of Internet of Things and Convergence
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• v.6 no.1
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• pp.1-8
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• 2020

Following the advanced fuel cycle initiative (AFCI) promotions in the United States, the analytic proposition for global fuel cycle initiative (GFCI) has been investigated using dynamical simulations. The political and economic aspects are considered simultaneously due to the particular characteristics of the nuclear materials. The spent nuclear fuels (SNFs) are treated as the reprocessing by the nuclear non-proliferation treaty (NPT) exemption nations and the NPT excluded nations. Otherwise, the pyroprocessing and repository can be done without NPT restriction. In addition, the international trade is considered as the economic aspect where the energy production is a key issue of the GFCI. The dynamical simulations have been done until 2050. The result of the International Trade shows the gradually increasing shape. Additionally, the Nuclear Power Plant Operation shows the increasing by stepwise shape.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.3
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• pp.191-197
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• 2017

The Korea Atomic Energy Research Institute (KAERI) has developed a safeguards technology for pyroprocessing based on the Safeguards-By-Design (SBD) concept. KAERI took part in a Member-State Support Program (MSSP) to establish a pyroprocessing safeguards approach. A Reference Engineering-scale Pyroprocessing Facility (REPF) concept was designed on which KAERI developed its safeguards system. Recently the REPF is being upgraded to the REPF+, a scaled-up facility. For assessment of the nuclear-material accountancy (NMA) system, KAERI has developed a simulation program named Pyroprocessing Material Flow and MUF Uncertainty Simulation (PYMUS). The PYMUS is currently being upgraded to include a Near-Real-Time Accountancy (NRTA) statistical analysis function. The Advanced Spent Fuel Conditioning Process Safeguards Neutron Counter (ASNC) has been updated as Non-Destructive Assay (NDA) equipment for input-material accountancy, and a Hybrid Induced-fission-based Pu-Accounting Instrument (HIPAI) has been developed for the NMA of uranium/transuranic (U/TRU) ingots. Currently, performance testing of Compton-suppressed Gamma-ray measurement, Laser-Induced Breakdown Spectroscopy (LIBS), and homogenization sampling are underway. These efforts will provide an essential basis for the realization of an advanced nuclear-fuel cycle in the ROK.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.261-267
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• 2010

The pyrometallurgical nuclear fuel recycle process, called pyroprocessing, has been known as a promising nuclear fuel recycling technology. Pyroprocessing technology is crucial to advanced nuclear systems due to increased nuclear proliferation resistance and economic efficiency. The basic concept of pyroprocessing is group actinide recovery, which enhances the nuclear proliferation resistance significantly. One of the key steps in pyroprocessing is "electrowinning" which recovers group actinides with lanthanide from the spent nuclear fuels. In this study, a vertical cadmium distiller was manufactured. The evaporation rate of pure cadmium in vertical cadmium distiller varied from 12.3 to $40.8g/cm^2/h$ within a temperature range of 773 923 K and pressure below 0.01 torr. Uranium - cadmium alloy was fabricated by electrolysis using liquid cadmium cathode in a high purity argon atmosphere glove box. The distillation behavior of pure cadmium and cadmium in uranium - cadmium alloy was investigated. The distillation behavior of cadmium from this study could be used to develop an actinide recovery process from a liquid cadmium cathode in a cadmium distiller.