Applications of genetic algorithms to optimization problems in the solvent extraction process for spent nuclear fuel

1997 ◽  
Vol 118 (1) ◽  
pp. 26-31 ◽  
Author(s):  
R. Omori
Keyword(s):  
Genetic Algorithms ◽  
Solvent Extraction ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Optimization Problems ◽  
Extraction Process

scholarly journals DEHBA (di-2-ethylhexylbutyramide) gamma radiolysis under spent nuclear fuel solvent extraction process conditions

2020 ◽  
Vol 170 ◽  
pp. 108608 ◽  
Author(s):  
Gregory P. Horne ◽  
Stephen P. Mezyk ◽  
Bruce J. Mincher ◽  
Christopher A. Zarzana ◽  
Cathy Rae ◽  
...  
Keyword(s):  
Solvent Extraction ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Extraction Process ◽  
Process Conditions ◽  
Gamma Radiolysis

scholarly journals New trends in the reprocessing of spent nuclear fuel. Separation of minor actinides by solvent extraction

2016 ◽  
Vol 71 (1) ◽  
pp. 123
Author(s):  
Jerzy Narbutt
Keyword(s):  
Solvent Extraction ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Natural Uranium ◽  
Heterocyclic Ligands ◽  
High Level Nuclear Waste ◽  
Extraction Processes ◽  
High Level

<p>Recycling of actinides from spent nuclear fuel by their selective separation followed by transmutation in fast reactors will optimize the use of natural uranium resources and minimize the long-term hazard from high-level nuclear waste. This paper describes solvent extraction processes recently developed, aimed at the separation of americium from lanthanide fission products as well as from curium present in the waste. Depicted are novel poly-N-heterocyclic ligands used as selective extractants of actinide ions from nitric acid solutions or as actinide-selective hydrophilic stripping agents.</p>

Study of nitrate complex formation with trivalent Pm, Eu, Am and Cm using a solvent extraction technique

2004 ◽  
Vol 92 (12) ◽  
Author(s):  
Sofie Andersson ◽  
C. Ekberg ◽  
J-O. Liljenzin ◽  
M. Nilsson ◽  
G. Skarnemark
Keyword(s):  
Solvent Extraction ◽  
Complex Formation ◽  
Nuclear Fuel ◽  
Stability Constants ◽  
Spent Nuclear Fuel ◽  
Aqueous Media ◽  
Separation Process ◽  
Nitrate Complex ◽  
The Stability ◽  
Solvent Extraction Technique

SummaryThe separation of actinides and lanthanides is an important question in the treatment of spent nuclear fuel in the transmutation concept. To find an efficient and well functioning separation process it is necessary to study the chemistry of the elements in the two groups in different aqueous media. The stability constants of the nitrate complex formation with Pm, Eu, Am and Cm were determined using solvent extraction. The extraction was studied using the synergistic system of 2,6-bis-(benzoxazolyl)-4-dodecyloxylpyridine and 2-bromodecanoic acid in

scholarly journals Dosimetry and methodology of gamma irradiation for degradation studies on solvent extraction systems

2020 ◽  
Vol 0 (0) ◽  
Author(s):  
Bart Verlinden ◽  
Peter Zsabka ◽  
Karen Van Hecke ◽  
Ken Verguts ◽  
Liviu-Cristian Mihailescu ◽  
...  
Keyword(s):  
Solvent Extraction ◽  
Ionizing Radiation ◽  
Gamma Irradiation ◽  
Dose Rate ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
High Dose ◽  
Spent Fuel ◽  
Irradiation Field ◽  
Pool Type

AbstractThe recycling of minor actinides from dissolved nuclear fuels by hydrometallurgical separation is one challenging strategy for the management of spent fuel. These future separation processes will likely be based on solvent extraction processes in which an organic solvent system (extractant and diluent) will be contacted with highly radioactive aqueous solutions. To establish a separation between different elements in spent nuclear fuel, many extractants have been studied in the past. A particular example is N,N,N′,N′-tetraoctyl diglycolamide (TODGA), which co-extracts lanthanides and actinides from nitric acid solutions into an organic phase (e.g. TODGA in n-dodecane). The radiolytic stability of these extractants is crucial, since they will absorb high doses of ionizing radiation during their usage. Worldwide, different gamma irradiation facilities are employed to expose extractants to ionizing radiation and gain insight in their radiation stability. The facilities differ in many ways, such as their environment (pool-type or dry), configuration and gamma sources (often 60Co or spent nuclear fuel). In this paper, a dosimetric assessment is made using different dosimeter systems in a pool-type irradiation facility, which has the advantage to be flexible in its arrangement of 60Co sources. It is shown that Red Perspex dosimeters can be used to accurately characterize this high dose rate gamma irradiation field (approx. 13.6 kGy h−1), after comparison with alanine, Fricke and ceric-cerous dosimetry in a lower dose rate gamma irradiation field (approx. 0.5 kGy h−1). A final validation of the whole chain of techniques is obtained by reproduction of the dose constants for TODGA in n-dodecane.

Formal transfer potentials of strontium and uranyl ions at water|1,2-dichloroethane interfaces

2012 ◽  
Vol 90 (10) ◽  
pp. 836-842 ◽  
Author(s):  
Tom J. Stockmann ◽  
Anne-Marie Montgomery ◽  
Zhifeng Ding
Keyword(s):  
Nuclear Fuel ◽  
Metal Ion ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Extraction Process ◽  
Nuclear Industry ◽  
Ion Transfer ◽  
Faradaic Current ◽  
Facilitated Ion Transfer ◽  
Transfer Potential

The extraction of dioxouranium (UO22+), or uranyl, and strontium (Sr2+) ions from spent nuclear fuel (SNF), often through a biphasic (aqueous / organic solvent) ligand assisted process, is critical for the implementation of a closed-loop nuclear fuel cycle whereby SNF is diverted from permanent geological disposal and the life of the nuclear industry is extended. Deeper understanding of the biphasic extraction process can be achieved through facile electrochemical experiments at a liquid|liquid interface. Of primary importance to developing a quantitative analysis of the ligand assisted or facilitated ion transfer (FIT) (i.e., transfer through interfacial complexation) case is to first quantify the free or simple metal IT; that is the amount of applied potential required to “push” ions across the water|organic interface. This value is, in fact, a constant referred to as the formal transfer potential ([Formula: see text]), which is unique to each metal ion in the biphasic system. Because of their hydrophilicity they often limit the polarizable potential window. Values for [Formula: see text], for the most part, have only been estimated. With a microinterface housed at the tip of a 25 µm capillary it is possible to reduce the Faradaic current to observe their transfer. Herein is described the quantification of [Formula: see text] and [Formula: see text] or the formal transfer potentials for dioxouranium and strontium ions, respectively.

scholarly journals Treatment of Spent Nuclear Fuel for Separation, Immobilization and Disposal of Heat Generating High Level Fission Products, Caesium and Strontium

2017 ◽  
Vol 30 (1) ◽  
pp. 37-42
Author(s):  
Md Akhlak Bin Aziz ◽  
Afrin Ahsan ◽  
Md Monsurul Islam Khan ◽  
Zahid Hasan Mahmood
Keyword(s):  
Fission Product ◽  
Nuclear Fuel ◽  
Nuclear Waste ◽  
Chemical Engineering ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Extraction Process ◽  
Safe Disposal ◽  
Synthetic Rock ◽  
High Level

Separation of heat generating, high level fission product caesium and strontium from spent nuclear fuel boosts the capacity of waste repositories by as much as fifty times. For efficient use of already scarce repositories, separation of such fission products is mandatory. Separations of caesium and strontium using Chlorinated Cobalt Dicarbollide (CCD), PEG (Polyethylene Glycol), UNEX process and by Calixarenes or Fission Product Extraction Process (FPEX) were discussed. The UNEX method was then proposed as the most feasible method option. Following separation, nuclear waste immobilization techniques for such high-level fission product were discussed. The techniques included usage of concrete, glass and synthetic rock. Among them synthetic rock was identified as the most suitable one for immobilization of high-level nuclear waste. Finally, a safe disposal system with necessary required geology was shown for safe disposal of the waste.Journal of Chemical Engineering, Vol. 30, No. 1, 2017: 37-42

Hafnium(IV) complexation with oxalate at variable temperatures

2017 ◽  
Vol 105 (5) ◽  
Author(s):  
Mitchell T. Friend ◽  
Nathalie A. Wall
Keyword(s):  
Solvent Extraction ◽  
Fission Product ◽  
Nuclear Fuel ◽  
Aqueous Phase ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Extraction Technique ◽  
Complexing Agents ◽  
Solvent Extraction Technique

AbstractAppropriate management of fission products in the reprocessing of spent nuclear fuel (SNF) is crucial in developing advanced reprocessing schemes. The addition of aqueous phase complexing agents can prevent the co-extraction of these fission products. A solvent extraction technique was used to study the complexation of Hf(IV) – an analog to fission product Zr(IV) – with oxalate at 15, 25, and 35°C in 1 M HClO

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