• Nuclear Engineering and Technology
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• v.43 no.4
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• pp.335-342
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• 2011

As part of the Department of Energy's Fuel Cycle Research and Development Program an electrochemical technology employing molten salts is being developed for recycle of metallic fast reactor fuel and treatment of light water reactor oxide fuel to produce a feed for fast reactors. This technology has been deployed for treatment of used fuel from the Experimental Breeder Reactor II (EBR-II) in the Fuel Conditioning Facility, located at the Materials and Fuel Complex of Idaho National Laboratory. This process is based on dry (non-aqueous) technologies that have been developed and demonstrated since the 1960s. These technologies offer potential advantages compared to traditional aqueous separations including: compactness, resistance to radiation effects, criticality control benefits, compatibility with advanced fuel types, and ability to produce low purity products. This paper will summarize the status of electrochemical development and demonstration activities with used nuclear fuel, including preparation of associated high-level waste forms.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.125-131
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• 2009

An electrolytic system (zinc anode-gallium cathode) was setup to evaluate the performance of several stirrers prepared for this study, where stirrers have been used to prevent uranium from forming dendrite on the cathode in pyrochemical process. In the case of no-stirring condition, zinc dendrites began to grow on the gallium surface in 1 hour and some dendrite grew out of the cathode crucible around 6 hours. When a rectangular stirrer or a tilt stirrer was rotated, at 40${\sim}$150 rpm, to mix the liquid gallium cathode, dendritic growth of zinc metal was prevented irrespective of revolution speed, but some of the deposits overflowed out of the cathode crucible owing to the large centrifugal forces at 150 rpm. The harrow stirrer did not nearly retard the dendrite growth at 40 rpm, but the dendrite growth was retarded at higher than 100 rpm and the zinc deposits also did not overflow at 150 rpm. Pounder could also prevent the dendrite growth to some extent but it had some difficulties in operation compared with other types of stirrers.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.92-106
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• 2005

The head-end processes of spent TRISO fuel have been reviewed to understand the current status and the limitations of the reported processes. The main concerns in the TRISO treatment are to effectively breach and separate the carbon and SiC layers composing the TRISO particles. The crush-bum scheme which was considered in the early stages of the development has been replaced by the crush-leach or $CO_2$ burning and the succeeding CO decomposition process because of a sequestration problem of $CO_2$ containing $^{14}C$. However there are still many obstacles to overcome in the reported processes. Hence, innovative thermomechanical and pyrochemical concepts to breach the coating layers of the TRISO particle with a minimized amount of second waste are proposed in this paper and their principles are described in detail.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.280-289
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• 2005

A pyrochemical process has been introduced and utilized so that the transmutation of spent PWR fuel in PEACER can produce mainly low and intermediate level waste for near surface disposal. Major radioactive nuclides from PEACER pyroprocessing are composed of TRU and LLFP. In this study, the requirement for the final waste from PEACER is evaluated based on the methodology for establishment of waste acceptance criteria. Also, sensitivity analysis for several input parameters is conducted in order to determine acceptable decontamination factor (DF) and LLFP removal efficiency and to find out input parameter that extremely have an effect on DE As a result of the study, LLFP removal efficiency, especially Sr-90 and Tc-99, is proved to be a major nuclide which contributes to annual dose by human intrusion scenario rather than TRU DF. More than $98.5\%$ of LLFP have to be removed to meet below dose constraint within the DF more than 5.0E+03. Besides, because of the relative short half-life of Sr-90, the increasing of the institutional control period is recommended for most important input parameter to determine DF.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.426-427
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• 2005

Advanced Spent Fuel Conditioning Process (ACP) is a pyrochemical process in which the spent fuel of PWR is transformed into the uranic metal ingot. Through this process, which has been developed in KAERI since 1998, the radioactivity, the radiotoxicity, the heat and the volume of the PWR spent fuel are reduced by a quarter of the original. To demonstrate a lab-scale process and extract the data for the later pilot-scale process, a demonstration facility of ACP (ACPF) is under construction and the lab-scale demonstration is slated for 2006. To establish the safeguardability of ACPF, a safeguards system including a neutron counter based on non-destructive assay, which is named as ACP Safeguards Neutron Counter (ASNC), the ACP Safeguards Surveillance System (ASSS) which consists of two neutron monitors and five IAEA cameras, and Laser Induced Breakdown System (LIBS) have been developed and are ready to be installed at ACPF. The target materials of ACP to assay with ASNC are categorized into three types among which the first is the uranic metal ingot, the second is the salt waste and the last is $UO_2$ and $U_{3}O_8$ powders, rod cuts and hulls. The Pu content of process nuclear materials can be accounted with ASNC. The ASSS is integrated in the ACP Intelligent Surveillance Software (AISS) in which the IAEA camera images and background signals at the rear doors of ACPF are displayed. The composition of special nuclear materials of ACP can be measured with LIBS which can be a supporting measurement tool for ASNC. The conceptual picture of safeguards system of ACPF is shown in Fig. 1.

• Nuclear Engineering and Technology
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• v.34 no.1
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• pp.42-59
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• 2002

In order to transmute long-lived radioactive nuclides such as transuranics(TRU), Tc-99, and I- l29 in LWR spent fuel, a preliminary conceptual design study has been performed for the accelerator driven subcritical reactor system, called HYPER(Hybrid Power Extraction Reactor) The core has a hybrid neutron energy spectrum: fast and thermal neutrons for the transmutation of TRU and fission products, respectively. TRU is loaded into the HYPER core as a TRU-Zr metal form because a metal type fuel has very good compatibility with the pyre- chemical process which retains the self-protection of transuranics at all times. On the other hand, Tc-99 and I-129 are loaded as pure technetium metal and sodium iodide, respectively. Pb-Bi is chosen as a primary coolant because Pb-Bi can be a good spallation target and produce a very hard neutron energy spectrum. As a result, the HYPER system does not have any independent spallation target system. 9Cr-2WVTa is used as a window material because an advanced ferritic/martensitic steel is known to have a good performance under a highly corrosive and radiation environment. The support ratios of the HYPER system are about 4∼5 for TRU, Tc-99, and I-129. Therefore, a radiologically clean nuclear power, i.e. zero net production of TRU, Tc-99 and I-129 can be achieved by combining 4 ∼5 LWRs with one HYPER system. In addition, the HYPER system, having good proliferation resistance and high nuclear waste transmutation capability, is believed to provide a breakthrough to the spent fuel problems the nuclear industry is faced with.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.965-974
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• 2022

Platinum anodes are widely used for metal oxides reduction in LiCl-Li₂O, however high-cost and low-corrosion resistance hinder their implementation. NiO-Li₂O ceramics is an alternative corrosion resistant anode material. Anode processes on platinum and NiO-Li₂O ceramics were studied in (80 mol.%) LiCl-(20mol.%)KCl and (80 mol.%)LiCl-(20 mol.%)KCl-Li₂O melts by cyclic voltammetry, potentiostatic and galvanostatic electrolysis. Experiments performed in the LiCl-KCl melt without Li₂O illustrate that a Pt anode dissolution causes the Pt²⁺ ions formation at 3.14 V and 550℃ and at 3.04 V and 650℃. A two-stage Pt oxidation was observed in the melts with the Li₂O at 2.40 ÷ 2.43 V, which resulted in the Li₂PtO₃ formation. Oxygen current efficiency of the Pt anode at 2.8 V and 650℃ reached about 96%. The anode process on the NiO-Li₂O electrode in the LiCl-KCl melt without Li₂O proceeds at the potentials more positive than 3.1 V and results in the electrochemical decomposition of ceramic electrode to NiO and O₂. Oxygen current efficiency on NiO-Li₂O is close to 100%. The NiO-Li₂O ceramic anode demonstrated good electrochemical characteristics during the galvanostatic electrolysis at 0.25 A/cm² for 35 h and may be successfully used for pyrochemical treating of spent nuclear fuel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.125-133
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• 2017

KAERI is conducting research on spent cladding hulls and additive metals to generate a solidification host matrix for the noble metal fission product waste in anode sludge from the electro-refining process to minimize the volume of waste that needs to be disposed of. In this study, alloy compositions Zr-17Cr, Zr-22Cr, and Zr-27Cr were prepared with or without eight noble metals representing fuel waste using induction melting. The microstructures of the resulting alloys were characterized and electrochemical corrosion tests were conducted to evaluate their corrosion characteristics. All the compositions had better corrosion characteristics than other Zr-based alloys that were evaluated for comparison. Analysis of the leach solution after the corrosion test of the Zr-22Cr-8NM specimen indicated that the noble metals were not leached during corrosion under 500 mV imposed voltage, which simulates a highly oxidizing disposal environment. The results of this study confirm that Zr-Cr based compositions will likely serve as chemically stable waste forms.

• Journal of the Korean Electrochemical Society
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• v.12 no.3
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• pp.276-281
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• 2009

Electrochemical behaviors of $U^{3+}$ and $Gd^{3+}$ were investigated in LiCl-KCl eutectic molten salt by using various electrochemical techniques. The electrodeposition and dissolution currents for uranium show the maximum at -1.51V and -1.35V, respectively while, for gadolinium,at -2.15V and -1.9V, respectively. In case of LiCl-KCl molten salt containing both of $U^{3+}$ and $Gd^{3+}$, the peak potential of electrodeposition of gadolinium shifts to more positive potential than in the solution without $U^{3+}$. The potentials in chronopotentiometric data suddenly dropped to negative value as soon as the reduction currents were applied and became constant at the potential around which the $U^{3+}$ and $Gd^{3+}$ are electrodeposited. The results of normal pulse voltammetry (NPV) and square wave voltammetry show that those methods can be used to qualitatively analyze the elements in the melts. Especially, the differentiation of NPV result was found to be useful for the separation of the peaks of which potentials are close each other.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.99-105
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• 2010

This paper investigated the characteristics of wasteform containing a spent zeolite used as a separating agent of FPs for recycling LiCl waste which would be generated from pyrochemical process of spent PWR fuel. In this study, a conventional borosilicate and Ca-rich glass were used as a consolidating agent for spent zeolite and it's mixing ratio was changed in the range, $25{\sim}35wt%$. The leach rates of Cs and Sr had about $0.1{\sim}0.01g/m^2day$ and $0.001{\sim}0.0001g/m^2day$, respectively. The leach resistance of Cs increased with amount of SAP and it showed about 10 times higher in the Ca-rich glass wasteform than in the conventional borosilciate glass wasteform. The compressive strength of wasteform was affected with the amount of glass. Thermal expansion rate of containing 30wt% glass has relatively lower than others. Also, the melting temperature was little changed in given mixing ratio of glass.