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8 pages, 1390 KiB

Open AccessArticle

DFT Prediction of Radiolytic Stability of Conformationally Flexible Ligands

by Anastasiia Smirnova, Maksim Yablonskiy, Vladimir Petrov and Artem Mitrofanov

Energies 2023, 16(1), 257; https://doi.org/10.3390/en16010257 - 26 Dec 2022

Cited by 5 | Viewed by 1155

Abstract

Radiolytic stability is one of the main requirements of the ligands for the reprocessing of spent nuclear fuel. The prediction of radiation stability based only on the 2D molecular structural formula allows us to accelerate and simplify the development of new ligands. Here, [...] Read more.

Radiolytic stability is one of the main requirements of the ligands for the reprocessing of spent nuclear fuel. The prediction of radiation stability based only on the 2D molecular structural formula allows us to accelerate and simplify the development of new ligands. Here, we used quantum chemistry to investigate the radiolytic behavior of water-soluble diglycolamides as one of the most popular ligands for spent nuclear fuel reprocessing. The accurate accounting of conformational mobility in the descriptors based on the Frontier Orbital Fukui theory allowed us to obtain a good correlation between theoretical and experimental data. Full article

(This article belongs to the Special Issue Treatment of Radioactive Waste and Sustainability Energy)

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17 pages, 3717 KiB

Open AccessArticle

Calculation of the Henry’s Constant and the Thickness of the Equilibrium Adsorption Layer of Radon in the Layer-by-Layer Measurement of the Sorbent Activity

by Eldar P. Magomedbekov, Aleksei O. Merkushkin, Veronika S. Pokalchuk, Alexander V. Obruchikov, Ilia Yu. Lukiyanchikov, Alexander S. Chepurnov and Elena A. Vanina

Energies 2022, 15(24), 9569; https://doi.org/10.3390/en15249569 - 16 Dec 2022

Cited by 1 | Viewed by 1194

Abstract

The radioactive gas radon is ubiquitous in the environment and is a major contributor to the human inhalation dose. It is the second leading cause of lung cancer after smoking. Radon concentrations are particularly high in the air of radon-hazardous facilities—uranium mines and [...] Read more.

The radioactive gas radon is ubiquitous in the environment and is a major contributor to the human inhalation dose. It is the second leading cause of lung cancer after smoking. Radon concentrations are particularly high in the air of radon-hazardous facilities—uranium mines and radioactive waste repositories containing radium. To reduce the dose load on the staff, air in these premises should be continuously or periodically purified of radon. Carbon adsorbers can be successfully used for this purpose. The design of sorption systems requires information on both equilibrium and kinetic parameters of radon dynamic adsorption. The traditional way of obtaining such characteristics of the sorbent is to analyze the breakthrough curves. The present paper proposes a simple alternative method for determining parameters of dynamic radon adsorption (Henry’s constant and equilibrium adsorption layer thickness) from the results of a layer-by-layer gamma-spectrometric measurement of the sorbent. The analytical equation for smooth distribution of radon activity in the sorbent layer is obtained based on equilibrium adsorption layer theory for elute chromatography (pulsed injection of radon into the column). Using the dynamic adsorption of 222Rn on AG-3 activated carbon as an example, both equilibrium (Henry’s constant) and kinetic (thickness of the equilibrium adsorption layer) parameters of the adsorption dynamics were calculated. It was shown that the exposure duration of the column bed in the air flow and superficial gas velocity do not affect the result of the Henry’s constant calculation. The dependence of the equilibrium adsorption layer thickness on the superficial gas velocity over a wide range of values (5–220 cm/min) is described by the van Deemter equation. It was shown that the optimum air flow velocity, which corresponds to the maximum effectiveness of the bed, is 15–30 cm/min. This corresponds to the minimum of the equilibrium adsorption layer thickness (about 0.6 cm). The developed mathematical model makes it easy to define both equilibrium and kinetic parameters of dynamic adsorption of radon based on discrete distribution of its activity over the sections of the adsorption column. These parameters can then be used to calculate and design gas delay systems. It can be useful for studying the sorption capacity of various materials relative to radon. Full article

(This article belongs to the Special Issue Treatment of Radioactive Waste and Sustainability Energy)

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24 pages, 6108 KiB

Open AccessArticle

Nuclear Melt Glass from Experimental Field, Semipalatinsk Test Site

by Irina E. Vlasova, Vasily O. Yapaskurt, Alexei A. Averin, Oleg E. Melnik, Denis A. Zolotov, Roman A. Senin, Tatiana R. Poliakova, Iurii M. Nevolin, Stepan N. Kalmykov and Andrey A. Shiryaev

Energies 2022, 15(23), 9121; https://doi.org/10.3390/en15239121 - 01 Dec 2022

Cited by 2 | Viewed by 1347

Abstract

Investigation of shocked materials provides unique information about behavior of substances in extreme thermodynamic conditions. Near surface nuclear tests have induced multiple transformations of affected soils. Examination of nuclear glasses and relics of entrapped minerals provides a unique database on their behavior under [...] Read more.

Investigation of shocked materials provides unique information about behavior of substances in extreme thermodynamic conditions. Near surface nuclear tests have induced multiple transformations of affected soils. Examination of nuclear glasses and relics of entrapped minerals provides a unique database on their behavior under an intense temperature flash. In this work, several types of nuclear fallout particles from historic tests at the Semipalatinsk test site are investigated using complementary analytical methods. Distribution of radionuclides in all types of samples is highly heterogeneous; domains with high content of radionuclides are often intermixed with non-active materials. There is no general correlation between chemical composition of the glassy matrix and content of radionuclides. In aerodynamic fallout, the main fraction of radionuclides is trapped in the outer glassy shell. Relics of quartz grains are always devoid of radionuclides, while glass regions of high activity have different composition. In contrast to underground tests, iron-rich minerals are not necessarily radioactive. In most cases, the glassy matrix in anhydrous and is strongly polymerized, and the Q³ silicate groups dominate. Temperature-induced transformations of entrapped minerals are discussed. Investigation of zircon grains shows absence of a direct correlation between degree of decomposition into constituting oxides, morphology of resulting baddeleyite, and maximum experienced temperature. For the first time, temperature history of a nuclear ground glass is estimated from Zr diffusion profiles from decomposing zircon grain. Full article

(This article belongs to the Special Issue Treatment of Radioactive Waste and Sustainability Energy)

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10 pages, 2311 KiB

Open AccessArticle

Cs Extraction from Chloride Media by Calixarene Crown-Ethers

by Marie Simonnet, Thomas Sittel, Patrik Weßling and Andreas Geist

Energies 2022, 15(20), 7724; https://doi.org/10.3390/en15207724 - 19 Oct 2022

Cited by 2 | Viewed by 1522

Abstract

Asse II salt mine, in Germany, contains low and intermediate-level radioactive waste that must be retrieved in the upcoming years. Potentially contaminated salts and brines will require treatment, with ¹³⁷Cs being the main contaminant. Cs⁺ is problematic to selectively recover due [...] Read more.

Asse II salt mine, in Germany, contains low and intermediate-level radioactive waste that must be retrieved in the upcoming years. Potentially contaminated salts and brines will require treatment, with ¹³⁷Cs being the main contaminant. Cs⁺ is problematic to selectively recover due to its chemical similarity with Na⁺ and K⁺ which are present in high quantities in a salt mine. This paper offers a novel solution for Cs⁺ separation from concentrated chloride salt media by solvent extraction with calixarene-crown-ether extractants in an alcoholic diluent. The proposed solvent extracts Cs⁺ at elevated chloride concentrations (3–4 M) while back-extraction is achieved by contacting the solvent with dilute (0.01 M) hydrochloric acid. Full article

(This article belongs to the Special Issue Treatment of Radioactive Waste and Sustainability Energy)

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12 pages, 2543 KiB

Open AccessArticle

Medium-Temperature Phosphate Glass Composite Material as a Matrix for the Immobilization of High-Level Waste Containing Volatile Radionuclides

by Anna V. Frolova, Sergey E. Vinokurov, Irina N. Gromyak and Sergey S. Danilov

Energies 2022, 15(20), 7506; https://doi.org/10.3390/en15207506 - 12 Oct 2022

Cited by 4 | Viewed by 1214

Abstract

The search for matrices and technological solutions for the reliable immobilization of volatile radionuclides and high-level waste (HLW) components is an actual radiochemical problem. Methods of obtaining of sodium alumino-iron phosphate (NAFP) and iron phosphate (FP) glass composite materials synthesized at temperatures of [...] Read more.

The search for matrices and technological solutions for the reliable immobilization of volatile radionuclides and high-level waste (HLW) components is an actual radiochemical problem. Methods of obtaining of sodium alumino-iron phosphate (NAFP) and iron phosphate (FP) glass composite materials synthesized at temperatures of 450–750 °C, their structure and hydrolytic stability were investigated in this paper. The structure of the samples was studied by XRD and SEM-EDS. It was shown that, in the case of FP materials, the phase composition varies depending on the synthesis temperature, while NAFP materials have a complex multiphase composition at all crystallization temperatures. It has been established that the samples of the obtained glass composite materials have a high hydrolytic stability. At the same time, FP material obtained at 650 °C are the most stable, which makes this medium-temperature method of synthesis promising for the immobilization of volatile HLW components. Full article

(This article belongs to the Special Issue Treatment of Radioactive Waste and Sustainability Energy)

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14 pages, 3052 KiB

Open AccessArticle

Use of Reduced Graphene Oxide to Modify Melamine and Polyurethane for the Removal of Organic and Oil Wastes

by Tamuna Bakhiia, Anna Yu. Romanchuk, Konstantin I. Maslakov, Alexey A. Averin and Stepan N. Kalmykov

Energies 2022, 15(19), 7371; https://doi.org/10.3390/en15197371 - 07 Oct 2022

Cited by 1 | Viewed by 1788

Abstract

Methods for obtaining efficient sorption materials based on highly porous melamine and polyurethane matrices modified with reduced graphene oxide were developed. These materials are promising for solving environmental problems such as water pollution with organic products by sorption treatment. Reduced graphene oxides (rGOs) [...] Read more.

Methods for obtaining efficient sorption materials based on highly porous melamine and polyurethane matrices modified with reduced graphene oxide were developed. These materials are promising for solving environmental problems such as water pollution with organic products by sorption treatment. Reduced graphene oxides (rGOs) were synthesized from graphene oxide suspensions using potassium hydroxide, ascorbic acid or hydrazine hydrate. Composites with obtained rGO and melamine and polyurethane foam were produced for further characterization. The composites demonstrate high sorption of organic pollutants (oil, diesel fuel and gasoline) and low sorption of water. The composites were comprehensively analyzed by physicochemical techniques (SEM, XPS, Raman spectroscopy, UV–Vis) to elucidate the mechanism of sorption. Full article

(This article belongs to the Special Issue Treatment of Radioactive Waste and Sustainability Energy)

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9 pages, 2813 KiB

Open AccessArticle

Synthesis of Mixed Actinide Oxides Using Microwave Radiation

by Konstantin Dvoeglazov, Yury Kulyako, Sergey Vinokurov, Boris Myasoedov, Mikhail Dmitriev, Oleg Ushakov, Yuri Mochalov, Andrei Shadrin and Pavel Smolkin

Energies 2022, 15(18), 6618; https://doi.org/10.3390/en15186618 - 09 Sep 2022

Cited by 4 | Viewed by 1346

Abstract

A method has been developed for producing mixed actinide oxides suitable for fabricating mixed nitride uranium plutonium fuel for fast neutron reactors. The method is based on the use of microwave radiation for the direct denitration of actinide nitrate solutions. The possibility of [...] Read more.

A method has been developed for producing mixed actinide oxides suitable for fabricating mixed nitride uranium plutonium fuel for fast neutron reactors. The method is based on the use of microwave radiation for the direct denitration of actinide nitrate solutions. The possibility of producing uranium, plutonium, and neptunium-mixed oxides was shown. A pilot installation for preparing actinide oxides by microwave denitration was designed and tested. Mixed oxides of uranium and cerium (for plutonium imitation) were successfully used to synthesize uranium cerium nitrides and produce fuel pellets. Compared with the precipitation (ammonia) method of producing mixed oxides, microwave denitration reduces the generation of secondary liquid radioactive waste by more than six times. Full article

(This article belongs to the Special Issue Treatment of Radioactive Waste and Sustainability Energy)

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13 pages, 4008 KiB

Open AccessArticle

Behavior of Glass-like and Mineral-like Phosphate Compounds with an Immobilized Chloride Mixture in Hydrogen Peroxide Solutions

by Anna V. Frolova, Svetlana A. Kulikova, Kseniya Y. Belova, Sergey S. Danilov and Sergey E. Vinokurov

Energies 2022, 15(17), 6477; https://doi.org/10.3390/en15176477 - 05 Sep 2022

Cited by 1 | Viewed by 1408

Abstract

A new type of high-level waste (HLW) is generated during pyrochemical reprocessing of mixed nitride spent uranium–plutonium nuclear fuel. Such waste is a spent electrolyte, which is a mixture of chloride salts containing approximately 25.7 wt.% LiCl + 31.6 wt.% KCl + 4.1 [...] Read more.

A new type of high-level waste (HLW) is generated during pyrochemical reprocessing of mixed nitride spent uranium–plutonium nuclear fuel. Such waste is a spent electrolyte, which is a mixture of chloride salts containing approximately 25.7 wt.% LiCl + 31.6 wt.% KCl + 4.1 wt.% CsCl + 5.1 wt.% BaCl₂ + 3.8 wt.% SrCl₂ + 29.7 wt.% LaCl₃, and its immobilization in reliable matrices is an actual radiochemical problem. The structure and hydrolytic stability of sodium aluminoironphosphate (NAFP) glass and a low-temperature mineral-like magnesium potassium phosphate (MPP) matrix, which are promising for spent electrolyte immobilization in the presence of hydrogen peroxide solutions simulating natural water radiolysis products, were studied in this work. The structure of the samples was studied using the SEM-EDS method. It was shown that the initial samples of NAFP glass after leaching in hydrogen peroxide solutions are prone to precipitation of crystalline phases on the surface, which are mainly represented by a mixture of sodium–iron–aluminum pyrophosphates. It was established that the leaching rate of structure-forming components of NAFP and MPP matrices generally increase, but remain at a low level, meeting modern requirements for HLW immobilization. This confirms the effectiveness of the studied matrices for the industrial use of the spent electrolyte. Full article

(This article belongs to the Special Issue Treatment of Radioactive Waste and Sustainability Energy)

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Review

Jump to: Research

35 pages, 9625 KiB

Open AccessReview

Application of Cement-Based Materials as a Component of an Engineered Barrier System at Geological Disposal Facilities for Radioactive Waste—A Review

by Ekaterina Aleksandrovna Tyupina, Pavel Pavlovich Kozlov and Victoria Valerievna Krupskaya

Energies 2023, 16(2), 605; https://doi.org/10.3390/en16020605 - 04 Jan 2023

Cited by 9 | Viewed by 2366

Abstract

Over the past several decades, the international community has been actively engaged in developing a safe method for isolating spent nuclear fuel, high and intermediate level radioactive wase of different degrees of heat generation in deep geological formations on the basis of regulatory [...] Read more.

Over the past several decades, the international community has been actively engaged in developing a safe method for isolating spent nuclear fuel, high and intermediate level radioactive wase of different degrees of heat generation in deep geological formations on the basis of regulatory requirements existing in each individual country (for example, in the Russian Federation-NP-055-14). Such a storage facility should be equipped with an engineered safety barrier system that combines a range of materials capable of ensuring the safe localization of environmentally and health-threatening nuclear power generation industry and the nuclear industry waste products, in particular. On the basis of the international experience discussed in this article on the design and operation of such facilities, the most universal material in terms of the functions performed as a component of the engineered barrier system is cement and the cement-based product mixed with various components—concrete. Furthermore, due to the possible mutual influence of buffer materials and their transformation over time at interfaces, this work considers the impact of cement-based barriers on other components of engineered barrier systems, the information on which has been accumulated as a result of both analytical laboratory tests and in situ radioactive waste disposal facilities under construction. Full article

(This article belongs to the Special Issue Treatment of Radioactive Waste and Sustainability Energy)

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23 pages, 3442 KiB

Open AccessReview

Approaches to Disposal of Nuclear Waste

by Michael I. Ojovan and Hans J. Steinmetz

Energies 2022, 15(20), 7804; https://doi.org/10.3390/en15207804 - 21 Oct 2022

Cited by 40 | Viewed by 10335

Abstract

We present a concise mini overview on the approaches to the disposal of nuclear waste currently used or deployed. The disposal of nuclear waste is the end point of nuclear waste management (NWM) activities and is the emplacement of waste in an appropriate [...] Read more.

We present a concise mini overview on the approaches to the disposal of nuclear waste currently used or deployed. The disposal of nuclear waste is the end point of nuclear waste management (NWM) activities and is the emplacement of waste in an appropriate facility without the intention to retrieve it. The IAEA has developed an internationally accepted classification scheme based on the end points of NWM, which is used as guidance. Retention times needed for safe isolation of waste radionuclides are estimated based on the radiotoxicity of nuclear waste. Disposal facilities usually rely on a multi-barrier defence system to isolate the waste from the biosphere, which comprises the natural geological barrier and the engineered barrier system. Disposal facilities could be of a trench type, vaults, tunnels, shafts, boreholes, or mined repositories. A graded approach relates the depth of the disposal facilities’ location with the level of hazard. Disposal practices demonstrate the reliability of nuclear waste disposal with minimal expected impacts on the environment and humans. Full article

(This article belongs to the Special Issue Treatment of Radioactive Waste and Sustainability Energy)

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