• Nuclear Engineering and Technology
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• 제55권3호
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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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• 제43권4호
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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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• 제43권4호
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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.

• 전기화학회지
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• 제12권3호
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• pp.276-281
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• 2009

본 연구는 고온화학공정에 사용할 전기화학적 센서 기술 개발의 사전 연구로서 고온 LiCl-KCl 공융염에 $UCl_3$나 $GdCl_3$를 녹여 $U^{3+}$와 $Gd^{3+}$의 전기화학 반응을 조사하였다. $U^{3+}$는 고온 LiCl-KCl 용융염내에서 -0.2V/-0.35 V에서 $U^{4+}$로의 산화/환원반응의, -1.51 V/-1.35 V에서 전착/해리 반응전류의 피크를 나타내었다. $Gd^{3+}$의 경우 -2.15 V에서 전착반응 피크를, -1.9 V에서 산화해리전류 피크를 나타내었다. $U^{3+}$와 $Gd^{3+}$의 혼합 고온 용융염에서는 $Gd^{3+}$의 전착 피크가 양의 전위로 이동하였다. 일정전류법의 결과는 시간이 지남에 따라 전위값이 일정해졌으며 그 전위값은 전해질 내 반응물의 환원전위값과 일치하였다. 노멀펄스전위법과 직각파전위법은 두 원소의 정성분석을 위한 좋은 전기화학적 방법론임을 보였으며 특히 노멀펄스전위곡선을 미분한 결과는 사용된 다른 방법에 비해 반응전류의 피크분리가 잘 일어났다.