Comparison of Closed Nuclear Fuel Cycle Technologies

Shadrin, A. Yu.; Ivanov, V. B.; Skupov, M. V.; Troyanov, V. M.; Zherebtsov, A. A.

doi:10.1007/s10512-016-0171-2

Cited by 7 publications

(2 citation statements)

References 5 publications

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“…Reprocessing with using different solvents. The chosen spent fuel reprocessing technologies: Molten chloride salts electrolysis, including actinides regeneration to metal on a solid-state cathode together with the water technology with organic extraction agents [10], [11]. The advantages of decentralized spent fuel reprocessing (within the SNFC) are:…”

Section: Fuel Reprocessing For "Brest" Reactorsmentioning

confidence: 99%

On-Site Nuclear Fuel Cycle of “BREST” Reactors

Glazov¹

2018

IJEPE

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Dynamic developing of modern nuclear industry demands meeting the following requirements: improved safety, reduced capital costs, radioactive waste (RW) management issues, independence of limited resources. Efficiency of uranium resources used in "BREST" reactors based on a closed fuel cycle is about 160 times higher than for VVER, RBMK reactors [1], which makes it possible to stop searching for new deposits and uranium mining. The need for periodical fuel regeneration and fabrication in a closed cycle includes reproduction of plutonium in the core without the uranium containing screens (breeding ratio is approximately 1,05, ensuring a high level of safety and support of the non-proliferation regime) and transmutation of the most dangerous long-lived actinides and high refining of RW, achieving the radiation balance of buried RW and extracted uranium ore. The manufacturing is located directly at the NPP to avoid transportation of fissile materials. This approach provides economic efficiency of the entire complex.

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Section: Fuel Reprocessing For "Brest" Reactorsmentioning

confidence: 99%

On-Site Nuclear Fuel Cycle of “BREST” Reactors

Glazov¹

2018

IJEPE

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“…На сегодняшний день в проекте «Прорыв» разрабатываются эффективные методы переработки облученного ядерного топлива (ОЯТ) для возвращения в топливный цикл долгоживущих радионуклидов с целью его замыкания [1][2][3]. Одной из задач разработки замкнутого топливного цикла является переработка высокоактивных азотнокислых рафинатов PUREXпроцесса [4][5][6]. Для реализации данной задачи требуется выделение наиболее радиоактивных минорных актинидов -америция-241 и кюрия-244, которые образуются при работе реактора и находятся в облученном топливе [5,7].…”

Section: Introductionunclassified

Salting out of americium-241 in the sorption process using a solid-phase extractant based on TODGA

Savelev,

Rachkov

2023

GNS

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Today, the «Proryv» project is developing effective methods of reprocessing irradiated nuclear fuel (SNF) to return long-lived radionuclides to the fuel cycle to close it. One of the challenges of closed fuel cycle development is the reprocessing of highly active nitric acid raffinates from the PUREX-process. To achieve this task, it is necessary to separate americium-241 from liquid radioactive waste. When processing and fractionating liquid radioactive waste, extraction and sorption technologies for the extraction, purification and concentration of radionuclides are widely used. The highest efficiency and selectivity in the extraction processes of actinoids (III) and lanthanides (III) with rare earth elements (REE) and transplutonium elements (TPE) from nitric acid solutions of spent nuclear materials reprocessing were shown by extractants based on N, N, N', N'-tetraoctyldiglycolamide (TODGA). Before using a solid-phase extractant based on TOGDA, the ions of the substance in solution must be converted to neutral complexes or other non-dissociated compounds. This can be achieved by adding neutral salts to the solution, which reduce the solubility of the elements to be separated, shift the extraction distribution and significantly increase the extraction efficiency. The extracted substance is extracted in the form of a new phase - solid precipitate, liquid or gas phase, and in the case of liquid extraction there is an increase in the capacity of the extractant for the target component. Therefore, the addition of salts-salting agents to the aqueous phase to increase the ionic strength of the solution increases the distribution coefficients of extracted substances, which in turn increases the capacity of sorbents. The purpose of this work is to study the process of salting out of americium-241 during sorption using an experimental modified sample of solid-phase extractant based on TODGA in the studied model solutions of liquid radioactive waste with a uranium macrocomponent for different NaNO3 contents. The study revealed that the highest distribution coefficients for the sorption of americium-241 and uranium were obtained in a solution containing 100 g/l NaNO3, but for uranium this effect is much less pronounced than for americium-241. During the study of the sorption kinetics of americium-241 and uranium, the salting effect was revealed, which is confirmed by the values of the equilibrium concentrations of americium-241 and uranium in solution at the same time point but with different NaNO3 concentrations. The difference in the equilibrium concentrations for americium-241 was an order of magnitude towards its decrease when NaNO3 concentration was increased up to 100 g/litre. The use of this effect makes it possible to obtain the maximum capacity for americium-241 in the system with uranium macrocomponents

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