• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1097-1104
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• 2023

A preliminary criticality analysis for novel pyrochemical apparatuses for the reprocessing of mixed uranium-plutonium nitride spent nuclear fuel from the BREST-OD-300 reactor was performed. High-temperature processing apparatuses, "metallization" electrolyzer, refinery remelting apparatus, refining electrolyzer, and "soft" chlorination apparatus are considered in this work. Computational models of apparatuses for two neutron radiation transport codes (MCU-FR and MCNP) were developed and calculations for criticality were completed using the Monte Carlo method. The criticality analysis was performed for different loads of fissile material into the apparatuses including overloading conditions. Various emergency situations were considered, in particular, those associated with water ingress into the chamber of the refinery remelting apparatus. It was revealed that for all the considered computational models nuclear safety rules are satisfied.

• Nuclear Engineering and Technology
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• v.43 no.4
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• pp.329-334
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• 2011

Two conceptual flowsheets were developed for recycling used nuclear fuel. One flowsheet was developed for recycling used oxide nuclear fuel from light water reactors while the other was developed for recycling used metal fuel from fast spectrum reactors. Both flowsheets were developed from a set of design principles including efficient actinide recovery, nonproliferation, waste minimization and commercial viability. Process chemistry is discussed for each unit operation in the flowsheet.

• 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 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.