Artículos de revistas: "Metallic nuclear waste"

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Stoulil, J. y D. Dobrev. "Microbial corrosion of metallic materials in a deep nuclear-waste repository". Koroze a ochrana materialu 60, n.º 2 (1 de junio de 2016): 59–67. http://dx.doi.org/10.1515/kom-2016-0010.

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AbstractThe study summarises current knowledge on microbial corrosion in a deep nuclear-waste repository. The first part evaluates the general impact of microbial activity on corrosion mechanisms. Especially, the impact of microbial metabolism on the environment and the impact of biofilms on the surface of structure materials were evaluated. The next part focuses on microbial corrosion in a deep nuclear-waste repository. The study aims to suggest the development of the repository environment and in that respect the viability of bacteria, depending on the probable conditions of the environment, such as humidity of bentonite, pressure in compact bentonite, the impact of ionizing radiation, etc. The last part is aimed at possible techniques for microbial corrosion mechanism monitoring in the conditions of a deep repository. Namely, electrochemical and microscopic techniques were discussed.

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Dietz, N. L. y D. D. Keiser. "TEM Analysis of Corrosion Products From a Radioactive Stainless Steel-based Alloy". Microscopy and Microanalysis 6, S2 (agosto de 2000): 368–69. http://dx.doi.org/10.1017/s1431927600034334.

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Argonne National Laboratory has developed an electrometallurgical treatment process for metallic spent nuclear fuel from the Experimental Breeder Reactor-II. This process stabilizes metallic sodium and separates usable uranium from fission products and transuranic elements that are contained in the fuel. The fission products and other waste constituents are placed into two waste forms: a ceramic waste form that contains the transuranic elements and active fission products such as Cs, Sr, I and the rare earth elements, and a metal alloy waste form composed primarily of stainless steel (SS), from claddings hulls and reactor hardware, and ∼15 wt.% Zr (from the U-Zr and U-Pu-Zr alloy fuels). The metal waste form (MWF) also contains noble metal fission products (Tc, Nb, Ru, Rh, Te, Ag, Pd, Mo) and minor amounts of actinides. Both waste forms are intended for eventual disposal in a geologic repository.

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Janney, D. E. y D. D. Keiser. "Actinides in metallic waste from electrometallurgical treatment of spent nuclear fuel". JOM 55, n.º 9 (septiembre de 2003): 59–60. http://dx.doi.org/10.1007/s11837-003-0032-z.

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4

Holt, Erika, Maria Oksa, Matti Nieminen, Abdesselam Abdelouas, Anthony Banford, Maxime Fournier, Thierry Mennecart y Ernst Niederleithinger. "Predisposal conditioning, treatment, and performance assessment of radioactive waste streams". EPJ Nuclear Sciences & Technologies 8 (2022): 40. http://dx.doi.org/10.1051/epjn/2022036.

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Before the final disposal of radioactive wastes, various processes can be implemented to optimise the waste form. This can include different chemical and physical treatments, such as thermal treatment for waste reduction, waste conditioning for homogenisation and waste immobilisation for stabilisation prior to packaging and interim storage. Ensuring the durability and safety of the waste matrices and packages through performance and condition assessment is important for waste owners, waste management organisations, regulators and wider stakeholder communities. Technical achievements and lessons learned from the THERAMIN and PREDIS projects focused on low- and intermediate-level waste handling is shared here. The recently completed project on Thermal Treatment for Radioactive Waste Minimization and Hazard Reduction (THERAMIN) made advances in demonstrating the feasibility of different thermal treatment techniques to reduce volume and immobilise different streams of radioactive waste (LILW) prior to disposal. The Pre-Disposal Management of Radioactive Waste (PREDIS) project addresses innovations in the treatment of metallic materials, liquid organic waste and solid organic waste, which can result from nuclear power plant operation, decommissioning and other industrial processes. The project also addresses digitalisation solutions for improved safety and efficiency in handling and assessing cemented-waste packages in extended interim surface storage.

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Barton, Daniel N. T., Thomas Johnson, Anne Callow, Thomas Carey, Sarah E. Bibby, Simon Watson, Dirk L. Engelberg y Clint A. Sharrad. "A review of contamination of metallic surfaces within aqueous nuclear waste streams". Progress in Nuclear Energy 159 (mayo de 2023): 104637. http://dx.doi.org/10.1016/j.pnucene.2023.104637.

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6

Moiseenko, V. y S. Chernitskiy. "Nuclear Fuel Cycle with Minimized Waste". Nuclear and Radiation Safety, n.º 1(81) (12 de marzo de 2019): 30–35. http://dx.doi.org/10.32918/nrs.2019.1(81).05.

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A uranium-based nuclear fuel and fuel cycle are proposed for energy production. The fuel composition is chosen so that during reactor operation the amount of each transuranic component remains unchanged since the production rate and nuclear reaction rate are balanced. In such a ‘balanced’ fuel only uranium-238 content has a tendency to decrease and, to be kept constant, must be sustained by continuous supply. The major fissionable component of the fuel is plutonium is chosen. This makes it possible to abandon the use of uranium-235, whose reserves are quickly exhausted. The spent nuclear fuel of such a reactor should be reprocessed and used again after separation of fission products and adding depleted uranium. This feature simplifies maintaining the closed nuclear fuel cycle and provides its periodicity. In the fuel balance calculations, nine isotopes of uranium, neptunium, plutonium and americium are used. This number of elements is not complete, but is quite sufficient for calculations which are used for conceptual analysis. For more detailed consideration, this set may be substantially expanded. The variation of the fuel composition depending on the reactor size is not too big. The model accounts for fission, neutron capture and decays. Using MCNPX numerical Monte-Carlo code, the neutron calculations are performed for the reactor of industrial nuclear power plant size with MOX fuel and for a small reactor with metallic fuel. The calculation results indicate that it is possible to achieve criticality of the reactor in both cases and that production and consuming rates are balanced for the transuranic fuel components. In this way, it can be assumed that transuranic elements will constantly return to such a reactor, and only fission products will be sent to storage. This will significantly reduce the radioactivity of spent nuclear fuel. It is important that the storage time for the fission products is much less than for the spent nuclear fuel, just about 300 years.

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7

Janney, Dawn E. "Host phases for actinides in simulated metallic waste forms". Journal of Nuclear Materials 323, n.º 1 (noviembre de 2003): 81–92. http://dx.doi.org/10.1016/j.jnucmat.2003.08.032.

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Rodríguez, Martín A. "Anticipated Degradation Modes of Metallic Engineered Barriers for High-Level Nuclear Waste Repositories". JOM 66, n.º 3 (1 de febrero de 2014): 503–25. http://dx.doi.org/10.1007/s11837-014-0873-7.

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9

Janney, D. E. "Incorporation of Actinide Elements into Iron-Zirconium Intermetallic Phases in Metallic Waste Forms for High-Level Nuclear Waste". Microscopy and Microanalysis 8, S02 (agosto de 2002): 1310–11. http://dx.doi.org/10.1017/s1431927602104983.

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10

Pavliuk, Alexander O., Evgeniy V. Bespala, Sergey G. Kotlyarevskiy, Ivan Yu Novoselov y Veleriy N. Kotov. "Analysis of Heat Release Processes inside Storage Facilities Containing Irradiated Nuclear Graphite". Science and Technology of Nuclear Installations 2022 (30 de enero de 2022): 1–13. http://dx.doi.org/10.1155/2022/2957310.

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The article is dedicated to the safety assessment of mixed storage of irradiated graphite and other types of radioactive waste accumulated during the operation of uranium-graphite reactors. The analysis of heat release processes inside storages containing irradiated nuclear graphite, representing a potential hazard due to the possible heating and, accordingly, the release of long-lived radionuclides during oxidation was carried out. The following factors were considered as the main factors that can lead to an increase in the temperature inside the storage facility: corrosion of metallic radioactive waste, the presence of fuel fragments, and also the random exposure of irradiated graphite to local sources of thermal energy (spark, etc.). It was noted in the work that the combined or separate influence of some factors can lead to an increase in the temperature of the onset of the initiation of Wigner energy release in graphite radwaste (Tin ≈ 90–100°C for the “Worst-case” graphite). The model of heat generation in the storage was developed based on the analysis of the features of graphite radioactive waste storage and Wigner energy release. The layered location of different types of waste (graphite and aluminum) and the local character of the distribution of heat sources were adopted in this model. The greatest heating is achieved if graphite radioactive waste is located near the concrete walls of the storage facility, as well as in direct contact with irradiated aluminum radioactive waste, which was shown in this paper.

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11

Pieraccini, Michel y Sylvain Granger. "A nuclear owner/operator perspective on ways and means for joint programming on predisposal activities". EPJ Nuclear Sciences & Technologies 6 (2020): 20. http://dx.doi.org/10.1051/epjn/2019039.

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Nuclear decommissioning is a worldwide competitive market. It is also the main source of radioactive waste from the nuclear energy field. In order to reduce the waste volume it is necessary to sort the actual radioactive waste to be disposed of and to separate them from other materials that could be recycled. Since 2015, Electricité de France (EDF) has gathered the waste management and dismantling (WM&D) projects, the related competences and human resources in the WM&D field, in a dedicated directorate (DP2D) and a company group called Cyclife (including waste treatment facilities). Taking into account the experience gained by carrying out its own WM&D projects as well as contributing to international cooperation, EDF considers that integrating collaborative research and development (R&D) on pre-disposal and waste management could be carried out following four main objectives: (1) alignment of the application of regulatory frameworks through appropriate definition of criteria and rules for radioactive waste to enable sensible worldwide comparison of technics; (2) improvement of technical and organisational aspects of nuclear reactors decommissioning using a demonstrator facility to be in operation, at first for graphite reactors, by 2022; (3) development of new techniques to decontaminate/homogenize metallic materials through a dedicated recycling route. These technics will be implemented in a new treatment facility foreseen to be available by 2030; and (4) increased training of decommissioning operators with the help of new technologies. All these improvements are aiming, beyond technical and experimental aspects, at reducing environmental impacts of nuclear activities as well as preserving the radioactive disposal volumes, as they are considered by EDF as rare resources.

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12

Galoisy, L., G. Calas, G. Morin, S. Pugnet y C. Fillet. "Structure of Pd–Te precipitates in a simulated high-level nuclear waste glass". Journal of Materials Research 13, n.º 5 (mayo de 1998): 1124–27. http://dx.doi.org/10.1557/jmr.1998.0158.

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Structural and bonding characteristics of simplified (Pd, Te) precipitates have been determined in a simulated nuclear French glass using extended x-ray absorption fine structure (EXAFS) and x-ray diffraction. In this sample, these precipitates have a homogeneous composition, with about 10 wt.% Te. They retain a face-centered cubic structure as in pure Pd with a cell parameter which obeys Vegard's law. Pd K-edge EXAFS shows the presence of Te in the Pd coordination shell, with (Pd–Te) distances of 2.80 Å. These distances, higher by 0.05 Å than the (Pd–Pd) distances, may result in a lower packing efficiency of the CFC lattice. The comparison with the average distances derived from x-ray diffraction shows the nonmetallic character of the Pd–Te bond in these precipitates. These bonding modifications may cause the limited solubility of Te in metallic Pd.

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13

Rébiscoul, Diane, Emilien Burger, Florence Bruguier, Nicole Godon, Jean-Louis Chouchan, Jean-Pierre Mestre, Pierre Frugier, Jean-Eric Lartigue y Stephane Gin. "Glass-Iron-Clay interactions in a radioactive waste geological disposal: a multiscale approach". MRS Proceedings 1518 (2013): 185–90. http://dx.doi.org/10.1557/opl.2013.67.

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ABSTRACTIn France, nuclear glass canisters arising from spent fuel reprocessing are expected to be disposed in a deep geological repository using a multi-barrier concept (glass/canister/steel overpack and claystone). In this context, glass - iron or corrosion products interactions were investigated in a clayey environment to better understand the mechanisms and driving forces controlling the glass alteration. Integrated experiments involving glass - metallic iron or magnetite - clay stacks were run at laboratory scale in anoxic conditions for two years. The interfaces were characterized by a multiscale approach using SEM, TEM, EDX and STXM at the SLS Synchrotron. Characterization of glass alteration patterns on cross sections revealed various morphologies or microstructures and an increase of the glass alteration with the proximity between the glass and the source of iron (metallic iron or magnetite) due to the consumption of the silica coming from the glass alteration. In case of magnetite, the silica consumption is mainly driven by a sorption of silica onto the magnetite. For experiments containing metallic iron, the silica consumption seems to be strongly driven by silicates precipitation including Fe and Fe/Mg when the Fe is not enough available. Moreover, in addition to Fe-silicates observed at the surface of the gel layers, iron is incorporated within the gel probably as nanosized precipitates of Fe-silicates which could affect its physical and chemical properties. Those results highlighted the impact of the distance between glass and iron source and the nature of the iron source which drive the process consuming the silica coming from the glass alteration.

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14

Mednikov, I. V., V. V. Vasilyev, A. S. Busygin y A. A. Sobko. "Provision of the radiation safety for the decomissioning of the heavy-water research nuclear reactor NRC «Kurchatov Institute» – ITEP". Radiatsionnaya Gygiena = Radiation Hygiene 13, n.º 1 (31 de marzo de 2020): 74–83. http://dx.doi.org/10.21514/1998-426x-2020-13-1-74-83.

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The article provides a brief description of organizational and technical measures aimed at ensuring radiation safety during the decommissioning of the heavy-water research nuclear reactor of Institute for Theoretical and Experimental Physics after A.I. Alikhanov of National Research Centre «Kurchatov Institute». Information is provided on the history and features of the operation of the reactor, including parameters and characteristics that are significant for planning and conducting work. The peculiarities of legal regulation in the field of ensuring radiation safety are given; regulatory acts and rules accompanying other activities during decommissioning and directly related to radiation safety are also considered. The paper describes the work done in preparation for dismantling, the initial and current state of the installation, forthcoming work with examples of dismantled equipment. Methods for handling radioactive waste arising during decommissioning are considered, including methods for fragmentation of large structural elements (examples of mechanical devices are given), methods for sorting according to different specific activity (high activity, low activity), radionuclide composition and physical properties (solid, metallic, non-metallic, liquid). A special method for handling liquid radioactive waste is described, which includes the collection and temporary storage system. To assess the radiation situation at workplaces during the dismantling of the reactor structures, calculations of radiation transfer were carried out on the running and shutdown reactor, during which it was established that the expected dose to the personnel when performing activities on decommissioning of TBR is much lower than the limit values, established by regulatory documents. In accordance with the estimated radiation doses, rules and instructions for personnel were determined, including the procedure for using personal protective equipment, the necessary measures for surface decontamination, etc. Information is given on the procedure for radiation monitoring at all stages of dismantling and at the final stages of decommissioning including control of premises, personnel, equipment, waste of various types, atmospheric air.

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Lemont, Florent, Patrice Charvin, Aldo Russello y Karine Poizot. "An Innovative Hybrid Process Involving Plasma in a Cold Crucible Melter Devoted to the Futur Intermediate Level Waste Treatment: The SHIVA Technology". Advances in Science and Technology 73 (octubre de 2010): 148–57. http://dx.doi.org/10.4028/www.scientific.net/ast.73.148.

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The cold crucible technology first developed for the treatment of the high level fission products can also be used for the direct treatment of intermediate level wastes. In this case, the wastes can be under the states of liquids or solids. The first experiments carried out for the direct treatment of ionic exchange resins emphasised the requirement of very high temperature on the surface of the glass. When this surface is to cold, the unperfected oxidation lead to a glass containing inclusion such as metallic compounds coming from the reduction of species contained in the waste. Thus, the quality of the glass could be not enough to meet with some specific requirements for long term storage. For few years, the Laboratory of the Innovative Processes has been studied the capability of a cold crucible to involved plasma torches ensuring the high temperature required for a complete oxidation of a large composition of waste. The developments and the assessment of different technological ways lead to build a cold crucible fitted with a bottom inductor together with twin plasma torches above the glass bath. This is the SHIVA process. The researches carried out on this innovative technology have shown the high efficiency of the combination for the treatment of a large variety of solid wastes. The oxidation is complete and the produced glass can be easily poured in a canister. This innovative process provides new perspective of treatment for a large variety of intermediate level waste stored on the ground of nuclear facilities.

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16

Gossé, S., C. Guéneau, S. Bordier, S. Schuller, A. Laplace y J. Rogez. "A Thermodynamic Approach to Predict the Metallic and Oxide Phases Precipitations in Nuclear Waste Glass Melts". Procedia Materials Science 7 (2014): 79–86. http://dx.doi.org/10.1016/j.mspro.2014.10.011.

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17

Niederleithinger, Ernst, Vera Lay, Christian Köpp, Erika Holt y Maria Oksa. "PREDIS: innovative ways for predisposal treatment and monitoring of low and medium radioactive waste". Safety of Nuclear Waste Disposal 1 (10 de noviembre de 2021): 9–10. http://dx.doi.org/10.5194/sand-1-9-2021.

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Abstract. The EURATOM PREDIS project (http://www.predis-h2020.eu, last access: 25 October 2021) targets the development and implementation of activities for predisposal treatment of radioactive waste streams other than nuclear fuel and high-level radioactive waste. It started on 1 September 2020 with a 4 year duration. The consortium includes 47 partners from 17 member states. The overall budget of the project is EUR 23.7 million, with EC contribution of EUR 14 million. The PREDIS project develops and increases the technological readiness level (TRL) of treatment and conditioning methodologies for wastes for which no adequate or industrially mature solutions are currently available, including metallic materials, liquid organic waste and solid organic waste. The PREDIS project also develops innovations in cemented waste handling and predisposal storage by testing and evaluating. The technical work packages align with priorities formulated within the Roadmap Theme 2 of EURAD (https://www.ejp-eurad.eu/sites/default/files/2021-09/2_Predisposal_Theme_Overview.pdf, last access: 15 October 2021), Nugenia Global Vision (https://snetp.eu/wp-content/uploads/2020/10/Global-vision-document-ves-1-april-2015-aa.pdf, last access: 15 October 2021) and with those identified by the project's industrial end users group (EUG). The PREDIS will produce tools guiding decision making on the added value of the developed technologies and their impact on the design, safety and economics of waste management and disposal. Four technical work packages are focusing on specific waste types: metallic, liquid organic, solid organic, and cemented wastes. For the first three, the main aim lies in processing, stabilizing, and packaging the different waste streams, e.g. by using novel geopolymers, to deliver items which are in line with national and international waste acceptance criteria. In contrast, the fourth technical work package has a different focus. To provide better ways for a safe and effective monitoring of cemented waste packages including prediction tools to assess the future integrity development during predisposal activities, several digital tools are evaluated and improved. Safety enhancement (e.g. less exposure of testing personnel) and cost-effectiveness are part of the intended impact. The work includes but is not limited to inspection methods, such as muon imaging, wireless sensors integrated into waste packages as well as external package and facility monitoring, such as remote fiber optic sensors. The sensors applied will go beyond radiation monitoring and include proxy parameters important for long-term integrity assessment (e.g. internal pressure). Sensors will also be made cost-effective to allow the installation of many more sensors compared to current practice. The measured data will be used in digital twins of the waste packages for specific simulations (geochemical, integrity) providing a prediction of future behavior. Machine learning techniques trained by the characterization of older waste packages will help to connect the models to the current data. All data (measured and simulated) will be collected in a joint database and connected to a decision framework to be used at actual facilities. The presentation includes detailed information about the various tools under consideration in the monitoring of cemented waste packages, their connection and first results of the research.

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18

Guillaume, D., A. Neaman, M. Cathelineau, R. Mosser-Ruck, C. Peiffert, M. Abdelmoula, J. Dubessy, F. Villiéras, A. Baronnet y N. Michau. "Experimental synthesis of chlorite from smectite at 300ºC in the presence of metallic Fe". Clay Minerals 38, n.º 3 (septiembre de 2003): 281–302. http://dx.doi.org/10.1180/0009855033830096.

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AbstractThe alteration and transformation behaviour of montmorillonite (bentonite from Wyoming, MX-80) in low-salinity solutions (NaCl, CaCl2) in the presence of metallic Fe (powder and 86461 mm plate) and magnetite powder was studied in batch experiments at 300ºC to simulate the mineralogical and chemical reactions of clays in contact with steel in a nuclear waste repository. The evolutions of pH and solution concentrations were measured over a period of 9 months. The mineralogical and chemical evolution of the clays was studied by XRD, SEM, Transmission Mössbauer Spectroscopy and TEM (EDS, HR imaging and EELS). Dissolution of the di-octahedral smectite of the starting bentonite was observed, in favour of newly formed clays (chlorite and saponite), quartz, feldspars and zeolite. The formation of Fe-chlorite was triggered by contact with the metallic Fe plate and Fe-Mg-chlorite at distance from the Fe plate (>2 mm).

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Orr, Robin, Hugh Godfrey, Chris Broan, Dave Goddard, Guy Woodhouse, Peter Durham, Andrew Diggle y John Bradshaw. "Formation and physical properties of uranium hydride under conditions relevant to metallic fuel and nuclear waste storage". Journal of Nuclear Materials 477 (agosto de 2016): 236–45. http://dx.doi.org/10.1016/j.jnucmat.2016.04.057.

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20

Paraskevoulakos, C., C. A. Stitt, K. R. Hallam, A. Banos, M. Leal Olloqui, C. P. Jones, G. Griffiths, A. M. Adamska, J. Jowsey y T. B. Scott. "Monitoring the degradation of nuclear waste packages induced by interior metallic corrosion using synchrotron X-ray tomography". Construction and Building Materials 215 (agosto de 2019): 90–103. http://dx.doi.org/10.1016/j.conbuildmat.2019.04.178.

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21

Buck, Edgar C., Alan L. Schemer-Kohrn y Jonathan B. Wierschke. "Technetium Incorporation into C14 and C15 Laves Intermetallic Phases". MRS Proceedings 1518 (2013): 117–22. http://dx.doi.org/10.1557/opl.2013.69.

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ABSTRACTLaves-type intermetallic phases have been observed to be the dominant phases in a series of alloy compositions being designed for the immobilization of technetium in a metallic waste form. The dominant metals in the alloy compositions were Fe-Mo and Fe-Mo-Zr. The alloy composition, Fe-Mo-Zr, also contained Pd, Zr, Cr, and Ni. Both non-radioactive rhenium-containing and radioactive technetium-bearing alloy compositions were investigated. In the Fe-Mo series, the phases observed were Fe2Mo (C14 Laves phase) and ferrite in agreement with predictions. Both Tc and Re resided predominantly in the Laves phases. In the Fe-Mo-Zr system, the phases included hexagonal C14 with the composition (Fe,Cr)2Mo, cubic C15 phase with a (Fe,Ni)2Zr composition, and the hcp phase Pd2Zr. The observation of these phases was in agreement with predictions. Re was found in the C14 intermetallic, (Fe,Cr)2Mo. Technetium was also observed to be partitioned preferentially into the (Fe,Cr)2Mo phase; however, this phase exhibited a cubic structure consistent with the C15 structural type. The composition of Laves phases is influenced by both the atomic size and electro-negativity of the constituent elements. The long-term release behavior of technetium under nuclear waste disposal conditions may be more dependent on the corrosion characteristics of these individual Laves phases containing Tc than the other metallic phases.

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TOKUHIRO, Tadashi, Joshua W. CAREY, Rolanda M. REED y Sita S. AKELLA. "SELECTIVE CAPTURE AND ENCAPSULATION OF METALLIC CATIONS BY HYDROGELS CONSISTING OF COPOLY(N-ISOPROPYLACRYLAMIDE/FUNCTIONAL MONOMER) NETWORKS". SOUTHERN BRAZILIAN JOURNAL OF CHEMISTRY 20, n.º 20 (20 de diciembre de 2012): 25–41. http://dx.doi.org/10.48141/sbjchem.v20.n20.2012.29_revista_2012a.pdf.

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Under the U.S. Department of Energy Waste Treatment Baseline and Integrated Waste Management Strategy the recycling of spent nuclear fuel to minimize waste, to assure maximum energy recovery, and to pursue science-based R&D to possibly eliminate the need for geologic waste repositories, are programmatic goals. We have developed both polymer gel and porous materials for the separation and adsorption of targeted contaminants. Here, we have investigated capture and encapsulation capabilities of hydrogels consisting of thermally-sensitive copoly[N-isopropylacrylamide(1-x) / functional monomer(x)] networks, where functional denotes carboxylic, hydroxyl, or cyanide group (mol fraction x); the captured and encapsulated species were: Cr3+, Co2+, Cu2+, Ni2+, Eu3+, Ho3+ and Tb3+ present in aqueous medium. Natural diffusions of cations into gel phase and the physico-chemical affinity of functional groups for cations played a major role in capturing cations. Encapsulation of cations trapped in hydrogels was achieved by loss of water and conformational transformation of networks through a volumetric phase transition. Experimental determinations of cation amounts (mass) and copolymer composition were carried out by atomic absorption and elemental analyses of carbon, nitrogen and hydrogen, respectively. We developed two approaches for determination of efficiency and selectivity metrics describing capture and encapsulation of cations by functional groups using two theories: 1) mean field theory and 2) first-order thermodynamic perturbation theory. The integrated results thus obtained show that: Cu2+ and Co2+ were selectively encapsulated by carboxylic and cyanide groups, respectively. Carboxylic and hydroxyl groups were superior extractants for Cr3+, Eu3+ and Ho3+. Further the cyanide group was also efiective for Eu3+ and Ho3+. However, all functional groups examined here were ineffective in capture and encapsulation ofNi2+.

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Chaudhry, Muhammad Junaid, Sascha Gentes, Alexander Heneka y Carla Olivia Krauß. "Wet sieving and magnetic separation for the treatment of radioactive secondary waste produced from waterjet abrasive suspension cutting". Safety of Nuclear Waste Disposal 2 (6 de septiembre de 2023): 9–10. http://dx.doi.org/10.5194/sand-2-9-2023.

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Abstract. Dismantling of reactor pressure vessels and their built-in components is an enormous challenge in the deconstruction of a nuclear power plant. Due to the years of exposure to neutron radiation, the activated components can only be dismantled and packaged remotely. For reasons of radiation protection, preference is given to techniques that can be used underwater due to the shielding effect. A cutting method that meets these requirements is the waterjet abrasive suspension cutting technique (WAS). The cutting tool is capable of slicing metallic internals and other materials using a pure jet of water mixed with an abrasive substance at high velocity and pressure. The process offers numerous technical advantages, but it has a major disadvantage in producing secondary waste. Due to the addition of the abrasive substance, the WAS process produces a waste mixture of inactive abrasive particles and radioactive steel particles (activated by neutron radiation) during the dismantling of steel components in nuclear facilities. Since the steel particles are radioactive when the reactor pressure vessel (RPV) and its internals are dismantled, this particle mixture currently has to be disposed of as radioactive waste. This leads to a doubling of the radioactive waste. Despite the technical advantages, the WAS process used for cutting purposes is a severely disadvantage from an economic point of view, considering the significant disposal costs of the radioactive waste. The research project NaMaSK (wet sieving and magnetic separation of grain mixtures to minimise secondary waste in the dismantling of nuclear facilities), funded by the German Federal Ministry of Education and Research (BMBF), aims to separate the two fractions (abrasive and steel particles) with the help of magnetic separation and wet sieving. For this purpose, a prototype separation system MaSK (magnetic separation of grain mixtures to minimise secondary waste in the dismantling of nuclear facilities) with a magnetic filter and sieve has already been built and tested, and it can separate up to 98 % of the steel particles from the mixture. The separation process aimed to reduce the total amount of secondary waste by reusing abrasive particles for further WAS cutting. In the new test plant NaMaSK, the mode of operation will be converted from a batch process to continuous operation to highlight the economic aspect of the separation process. In this regard, an efficient design of the sieve structure and magnetic filter, followed by process optimisation, will be implemented. These new developments and the first results will be presented at the conference.

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Fujita, Reiko, Mitsuaki Yamaoka, Masatoshi Kawashima, Masaki Saito, Haruaki Matsuura y Hiroshi Akatsuka. "A metallic fuel cycle for Self-Consistent Nuclear Energy System (SCNES)". Progress in Nuclear Energy 40, n.º 3-4 (abril de 2002): 615–20. http://dx.doi.org/10.1016/s0149-1970(02)00057-4.

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Guo, Xiaolei, Penghui Lei, Chandi Mohanty, Tiankai Yao, Jie Lian y Gerald S. Frankel. "Enhanced Crevice Corrosion of Stainless Steel 316 By Degradation of Cr-Containing Hollandite Crevice Former". ECS Meeting Abstracts MA2022-02, n.º 11 (9 de octubre de 2022): 739. http://dx.doi.org/10.1149/ma2022-0211739mtgabs.

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The disposal of high-level nuclear waste (HLW) is an important societal problem. It is also an extremely challenging corrosion research topic due to the highly complicated waste chemistry, long lasting radiation effect, and inevitable exposure to aqueous environments for hundreds of thousands of years. To safely accommodate some critical radionuclides such as volatile 129I and strongly heat generating 137Cs, numerous crystalline ceramic waste forms have been created and studied. Among the promising nuclear waste forms, hollandite was developed to isolate Cs waste and the corresponding decay product Ba due to the open tunnel structure capable of accommodating mono- or divalent cations. In this study, the synergistic corrosion interactions between a Cr-containing hollandite (Ba1.15Cr2.3Ti5.7O16) and stainless steel (SS) 316 is explored. This is relevant to the permanent disposal of HLW involving the encasement of glass or crystalline ceramic waste forms containing immobilized radionuclides in a metallic canister made from corrosion resistant alloys such as SS. The experiments performed in this study simulate the potential corrosion interactions occurring at the interface of the SS and ceramic. After corroding the SS316 and Cr-containing hollandite in proximity in 0.6 M NaCl at 90 oC for 28 days, severe crevice corrosion was identified on the surface of the SS, as evidenced by the presence of large pits and crevice damage with diameters of hundreds of microns. Similarly, localized damage was also found at matching sites on the Cr-hollandite surface, indicating interactions between the two materials. The synergistic corrosion interaction was likely due to the continuous release of Cr3+ cations from both materials, including the passive dissolution of the SS and the heterogenous degradation of the Cr-hollandite. The extra Cr3+ cations originated from the hollandite reduces the incubation time required to reach the critical crevice solution condition, thereby accelerating the breakdown of the passive film of SS and the subsequent onset of active dissolution. An adapted crevice corrosion model is applied to quantitively explain the accelerated corrosion of SS observed in this study.

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Osacký, Marek, Miroslav Honty, Jana Madejová, Thomas Bakas y Vladimír Šucha. "Experimental interactions of Slovak bentonites with metallic iron". Geologica Carpathica 60, n.º 6 (1 de diciembre de 2009): 535–43. http://dx.doi.org/10.2478/v10096-009-0039-7.

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Experimental interactions of Slovak bentonites with metallic ironThe experimental stability of four bentonites and one K-bentonite from Slovak deposits in the presence of iron was studied to simulate the possible reactions of clays (bentonite barrier) in the contact with Fe containers in a nuclear waste repository. The batch experiments were performed at 60 °C for 30 and 120 days in aerobic conditions. The reaction products were examined by XRD, FTIR, and Mössbauer spectroscopies and CEC (cation exchange capacities) were determined. Reaction solutions were analysed for selected elements using AAS (atomic absorption spectrometry). The results show that bentonites do not interact equally with metallic iron. Bentonites from the Jelšový Potok, Kopernica and Lieskovec deposits reacted similarly whereas the interaction between the bentonite from Lastovce and the iron was less intensive. The lower reactivity of the bentonite from Lastovce can be explained by its low content of smectite. During iron-clay interactions the iron was consumed and Fe oxides (magnetite, lepidocrocite) were formed. Decrease of the smectite diffraction peaks intensity and CEC values during the experiments show rather the rearrangement of the original smectite crystals than dissolution of smectite. In the K-bentonite from the Dolná Ves deposit where the mixed-layer illite-smectite is present instead of smectite, the dissolution of illite-smectite was observed along with the neoformation of smectite. The structure of illite-smectite deteriorated more than the structure of smectites which suggests that this mixed-layer illite-smectite is much less stable in the presence of iron than smectites.

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27

Pokhitonov, Yu A., V. A. Starchenko, I. Yu Dalyaev y S. L. Titov. "Using hot isostatic pressing for radioactive waste isolation purposes". Radioactive Waste 16, n.º 3 (2021): 20–29. http://dx.doi.org/10.25283/2587-9707-2021-3-20-29.

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The paper summarizes the findings of a study focused on hot isostatic pressing (HIP) technique implemented by the Khlopin Radium institute. The equipment was designed and manufactured at the Kharkov’s Physics and Technology Institute. The installation provided a pressure of up to 400 NPa with the pressing temperature of up to 1250°C. The experiments were carried out on installations located in hot cells in the radiochemical department (Gatchina city). Samples of materials for HLW immobilization (titanate ceramics of the synroc type, stabilized cubic zirconia) and matrices for 129I immobilization based on copper iodide and metallic copper were obtained. The leaching rate from these samples of HLW elements (simulators) amounted to (0.5—1.5)·10–9 g/(cm2 ·day). Despite the high-performance characteristics of the materials obtained, some problems were revealed associated with the remote maintenance of equipment and the lack of industrial design analogues. Considering the experience gained, we believe that fairly simple equipment can be designed implying no complex systems and providing minimum preparatory operations. Joint efforts of technologists and designers will enable the automatization of equipment management and control through local control systems. Material loading and unloading operations can be robotized as well. Such technical solutions are expected to be in demand at industrial facilities for HLW final disposal (or when handling damaged fuel during the decommissioning of radiation and nuclear hazardous facilities).

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28

Haas, Allan, Dale F. Rucker y Marc T. Levitt. "Investigating the effective resistivity of reinforced concrete waste storage tanks at the Hanford Site". GEOPHYSICS 87, n.º 1 (19 de noviembre de 2021): B31—B43. http://dx.doi.org/10.1190/geo2021-0187.1.

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Industrialized sites pose challenges for conducting electrical resistivity geophysical surveys because the sites typically contain metallic infrastructure that can mask electrolytic-based soil and groundwater contamination. The Hanford Nuclear Site in eastern Washington State, USA, is an industrialized site with underground storage tanks, piping networks, steel fencing, and other potentially interfering infrastructure that could inhibit the effectiveness of electrical resistivity tomography (ERT) to map historical and monitor current waste releases. The underground storage tanks are the largest contributor by volume to subsurface infrastructure and can be classified as reinforced concrete structures with an internal steel liner. Directly measuring the effective value for the electrical resistivity of the tanks, that is, the combination of individual components that comprise the tank’s shell, is not reasonably possible because they are buried and are dangerously radioactive. Therefore, we indirectly assess the general resistivity of the tanks and the surrounding infrastructure by developing synthetic ERT models with a parametric forward-modeling study using a wide range of resistivity values from [Formula: see text] to [Formula: see text], which are equivalent to steel and dry rock, respectively. The synthetic models use the long-electrode ERT (LE-ERT) method, whereby steel-cased metallic wells surrounding the tanks are used as electrodes. The patterns and values of the synthetic tomographic models are then compared with LE-ERT field data from the AX tank farm at the Hanford Site. This indirect method of assessing the effective resistivity reveals that the reinforced concrete tanks are electrically resistive and the accompanying piping infrastructure has little influence on the overall resistivity distribution when using electrically based geophysical methods for characterizing or monitoring waste releases. Our findings are consistent with nondestructive testing literature that also indicates reinforced concrete to be generally resistive.

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29

Khorasanov, Georgiy, Dmitriy Samokhin, Aleksandr Zevyakin, Yevgeniy Zemskov y Anatoliy Blokhin. "Lead reactor of small power with metallic fuel". Nuclear Energy and Technology 4, n.º 2 (26 de noviembre de 2018): 99–102. http://dx.doi.org/10.3897/nucet.4.30527.

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The possibility for obtaining a hard neutron spectrum in small reactor cores has been considered. A harder spectrum than spectra in known fast sodium cooled and molten salt reactors has been obtained thanks to the selection of relatively small core dimensions and the use of metallic fuel and natural lead (natPb) coolant. The calculations for these compositions achieve an increased average neutron energy and a large fraction of hard neutrons in the spectrum (with energies greater than 0.8 MeV) caused by a minor inelastic interaction of neutrons with the fuel with no light chemical elements and with the coolant containing 52.3% of 208Pb, a low neutron-moderating isotope. An interest in creating reactors with a hard neutron spectrum is explained by the fact that such reactors can be practically used as special burners of minor actinides (MA), and as isotope production and research reactors with new consumer properties. With uranium oxide fuel (UO2) substituted by metallic uranium-plutonium fuel (U-Pu-Zr), the reactors under consideration have the average energy of neutrons and the fraction of hard neutrons increasing from 0.554 to 0.724 MeV and from 18 to 28% respectively. At the same time, the one-group fission cross-section of 241Am increases from 0.359 to 0.536 barn, while the probability of the 241Am fission increases from 22 to 39%. It is proposed that power-grade plutonium resulting from regeneration of irradiated fuel from fast sodium cooled power reactors be used as part of the fuel for future burner reactors. It contains unburnt plutonium isotopes and some 1% of MAs which transmutate into fission products in the process of being reburnt in a harder spectrum. This will make it possible to reduce the MA content in the burner reactor spent fuel and to facilitate so the long-term storage conditions for high-level nuclear waste in dedicated devices.

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30

Voronkova, Liubov V. "Recent Advances in Cast Iron Structure and Properties Ultrasonic Testing and Flaw Detection". Materials Science Forum 925 (junio de 2018): 499–503. http://dx.doi.org/10.4028/www.scientific.net/msf.925.499.

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The advantage of ultrasonic testing of iron castings prior to destructive control consists of an impressive reduction in time and costs, as well as the ability to assess the quality of the metal anywhere on the castings. According to the acoustic characteristics of cast iron it is possible to determine the form of graphite in it and its strength. The presence of chill in the metallic base and the thickness of the chilled layer is also determined by ultrasonic method. The use of electronic signal processing allows to distinguish it from high structural noise, which makes possible the testing of cast iron for a thickness of several meters. The use of transducers with phased array is the basis of the flaw detection of containers for nuclear waste from cast iron with globular graphite with a thickness of 500 mm.

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Grigaliūnienė, Dalia, Robertas Poškas, Raimondas Kilda, Hussam Jouhara y Povilas Poškas. "Modeling radionuclide migration from activated metallic waste disposal in a generic geological repository in Lithuania". Nuclear Engineering and Design 370 (diciembre de 2020): 110885. http://dx.doi.org/10.1016/j.nucengdes.2020.110885.

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32

Siddiqui, N. A., M. A. Iqbal, H. Abbas y D. K. Paul. "Reliability analysis of nuclear containment without metallic liners against jet aircraft crash". Nuclear Engineering and Design 224, n.º 1 (septiembre de 2003): 11–21. http://dx.doi.org/10.1016/s0029-5493(03)00080-3.

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33

Gebauer, Jana, Florian Gruber, Wilhelm Holfeld, Wulf Grählert y Andrés Fabián Lasagni. "Prediction of the Quality of Thermally Sprayed Copper Coatings on Laser-Structured CFRP Surfaces Using Hyperspectral Imaging". Photonics 9, n.º 7 (21 de junio de 2022): 439. http://dx.doi.org/10.3390/photonics9070439.

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With the progressive replacement of metallic parts by high-performance fiber-reinforced plastic (FRP) components, typical properties of metals are required to be placed on the material’s surface. A metallic coating applied to the FRP surface by thermal spraying, for instance, can fulfill these requirements, including electrical conductivity. In this work, laser pre-treatments are utilized for increasing the bond strength of metallic coatings. However, due to the high-precision material removal using pulsed laser radiation, the production-related heterogeneous fiber distribution in FRP leads to variations in the structuring result and consequently to different qualities of the subsequent coating. In this study, hyperspectral imaging (HSI) technologies in conjunction with deep learning were applied to carbon fiber-reinforced plastics (CFRP) structured by nanosecond pulsed laser. HSI-based prediction models could be developed, which allow for reliable prediction, with an accuracy of around 80%, of which laser-treated areas will successfully be coated and which will not. By using this objective and automatic evaluation, it is possible to avoid large amounts of rejects before further processing the parts and also to optimize the adhesion of coatings. Spatially resolved data enables local reworking during the laser process, making it feasible for the manufacturing process to achieve zero waste.

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34

Rucker, Dale F., Meng H. Loke, Marc T. Levitt y Gillian E. Noonan. "Electrical-resistivity characterization of an industrial site using long electrodes". GEOPHYSICS 75, n.º 4 (julio de 2010): WA95—WA104. http://dx.doi.org/10.1190/1.3464806.

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An electrical-resistivity survey was completed at the T tank farm at the Hanford nuclear site in Washington State, U.S.A. The purpose of the survey was to define the lateral extent of waste plumes in the vadose zone in and around the tank farm. The T tank farm consists of single-shell tanks that historically have leaked and many liquid-waste-disposal facilities that provide a good target for resistivity mapping. Given that the site is highly industrialized with near-surface metallic infrastructure that potentially could mask any interpretable waste plume, it was necessary to use the many wells around the site as long electrodes. To accommodate the long electrodes and to simulate the effects of a linear conductor, the resistivity inversion code was modified to assign low-resistivity values to the well’s location. The forward model within the resistivity code was benchmarked for accuracy against an analytic solution, and the inverse model was tested for its ability to recreate images of a hypothetical target. The results of the tank-farm field survey showed large, low-resistivity targets beneath the disposal areas that coincided with the conceptual hydrogeologic models developed regarding the releases. Additionally, in areas of minimal infrastructure, the long-electrode method matched the lateral footprint of a 3D surface-resistivity survey with reasonable fidelity. Based on these results, the long-electrode resistivity method may provide a new strategy for environmental characterization at highly industrialized sites, provided a sufficient number and density of wells exist.

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35

Parsons, Sam, Thomas Haines, Matthew White, Simon Norris, Johan Bertrand, Lise Griffault, Oliver Hall, Patrik Sellin y Christopher Harbord. "Features, events and processes (FEP) analysis of the interactions between repository monitoring systems and multi-barrier systems". Safety of Nuclear Waste Disposal 2 (6 de septiembre de 2023): 177–78. http://dx.doi.org/10.5194/sand-2-177-2023.

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Abstract. In line with the European waste directive 2011/70/EURATOM, the European Joint Programme on Radioactive Waste Management (EURAD) was launched in June 2019. EURAD aims to make a step change in European collaboration between advanced and early-stage waste management programmes. Monitoring Equipment and Data Treatment for Safe Repository Operation and Staged Closure (MODATS) is Work Package (WP) 17 of the EURAD programme. This WP aims to address a range of research needs relating to repository monitoring data and technologies. This abstract summarises research relating to the interactions between repository monitoring systems and the multi-barrier systems in which they are emplaced. A detailed features, events and processes (FEP) analysis is being undertaken to identify and describe the manner in which monitoring systems may interact with multi-barrier systems. Monitoring system components are generically considered in this research; they include data acquisition technologies, power technologies, data transmission technologies and data loggers. No specific disposal concepts have been selected; however, a range of commonly engineered barrier system materials and typical host rocks are considered. This FEP analysis research will provide an understanding of the impacts of monitoring systems, which could be used to aid monitoring system design optimisation, for example by supporting the screening of monitoring technologies (White and Scourfield, 2019). It may also provide evidence to demonstrate that monitoring systems do not unacceptably impact the safety functions of the multi-barrier system. Potential interactions may include the creation of gas migration pathways along monitoring power or data transmission cables positioned in the multi-barrier system. This process has been observed in underground research laboratory experiments, such as the Large Scale Gas Injection Test (LASGIT) experiment in the Äspö Hard Rock Laboratory, which involved a series of gas injection tests in a full-scale KBS-3V deposition hole. Other possible interactions could include (non-exhaustive) corrosion of metallic components, degradation of non-metallic components, microbial activity, gas generation, void introduction and formation and volume changes. The final output of this research will be a catalogue of FEPs describing the interactions between generic monitoring systems and multi-barrier systems. Similar to the Nuclear Energy Agency FEP lists, this FEP catalogue is intended to be a starting reference for waste management organisations to use to understand the potential interactions between their specific monitoring systems and multi-barrier systems.

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36

Law, Kathleen A., Stephen Parry, Nicholas D. Bryan, Sarah L. Heath, Steven M. Heald, Darrell Knight, Luke O’Brien et al. "Plutonium Migration during the Leaching of Cemented Radioactive Waste Sludges". Geosciences 9, n.º 1 (8 de enero de 2019): 31. http://dx.doi.org/10.3390/geosciences9010031.

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One of the most challenging components of the UK nuclear legacy is Magnox sludge, arising from the corrosion of Mg alloy-clad irradiated metallic U fuel that has been stored in high pH ponds. The sludges mainly comprise Mg hydroxide and carbonate phases, contaminated with fission products and actinides, including Pu. Cementation and deep geological disposal is one option for the long-term management of this material, but there is a need to understand how Pu may be leached from the waste, if it is exposed to groundwater. Here, we show that cemented Mg(OH)2 powder prepared with Pu(IV)aq is altered on contact with water to produce a visibly altered ‘leached zone’, which penetrates several hundred microns into the sample. In turn, this zone shows slow leaching of Pu, with long-term leaching rates between 1.8–4.4 × 10−5% of total Pu per day. Synchrotron micro-focus X-ray fluorescence mapping identified decreased Pu concentration within the ‘leached zone’. A comparison of micro-focus X-ray absorption spectroscopy (µ-XAS) spectra collected across both leached and unleached samples showed little variation, and indicated that Pu was present in a similar oxidation state and coordination environment. Fitting of the XANES spectra between single oxidation state standards and EXAFS modeling showed that Pu was present as a mixture of Pu(IV) and Pu(V). The change in Pu oxidation from the stock solution suggests that partial Pu oxidation occurred during sample ageing. Similarity in the XAS spectra from all samples, with different local chemistries, indicated that the Pu oxidation state was not perturbed by macro-scale variations in cement chemistry, surface oxidation, sample aging, or the leaching treatment. These experiments have demonstrated the potential for leaching of Pu from cementitious waste forms, and its underlying significance requires further investigation.

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37

Paraskevoulakos, C., K. R. Hallam y T. B. Scott. "Grout durability within miniaturised Intermediate Level Waste drums at early stages of interior volume expansion induced by encapsulated metallic corrosion". Journal of Nuclear Materials 510 (noviembre de 2018): 348–59. http://dx.doi.org/10.1016/j.jnucmat.2018.08.028.

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38

Galliez, Kévin, Guillaume Jossens, Alain Godot y Christophe Mathonat. "Characterization of Low Level Wastes: a new design for calorimetric measurement". EPJ Web of Conferences 170 (2018): 07003. http://dx.doi.org/10.1051/epjconf/201817007003.

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Calorimetry is one of the best solutions to estimate the overall quantity of nuclear material on a wide range of masses, from a few milligrams up to kilograms of radionuclides, by measuring the overall thermal power due to the radioactive decay coming from the waste contained in a metallic drum or a different type of container. It has many advantages as it features a non-destructive method which remains independent of matrix effect or the chemical composition. Until now, calorimetry allows to measure at the lowest 0.5 to 1 mW for samples up to 385 liters. But nowadays, thanks to new technological breakthroughs, KEP-Technologies calorimeters are able to measure as low as 50 μW for 40 liters samples. The μLVC is based on a new design with twin cells, a new temperature regulation loop and a heat-flow measurement system inside a vacuum chamber (Patent deposit P005299 LA/VL). The μLVC is a differential heat-flow calorimeter for precise measurement independent of the residual fluctuations caused by environmental changes. The new calorimeter is an industrial product able to work in environmental conditions with wide temperature variations. The first results have shown a great improvement in the detection of very low thermal effect thanks to the thermal noise reduction. The paper presents the developments in Large Volume Calorimetry as a new tool for quantification of nuclear material to characterize Pu-Am samples, i-graphite, and low tritium samples with high precision and reliability.

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39

Kuzina, Yu, D. Klinov, G. Mikhailov, A. Sorokin y V. Alekseev. "COMPLEX OF EXPERIMENTAL FACILITIES FOR DESIGN AND SAFETY JUSTIFICATION OF FAST REACTORS WITH LIQUID METAL COOLANTS". PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2021, n.º 4 (26 de diciembre de 2021): 172–94. http://dx.doi.org/10.55176/2414-1038-2021-4-172-194.

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To substantiate the safety and characteristics of fast reactors with liquid metal coolants, a complex of more than 20 stands of various profiles and purposes, well equipped with modern measuring instruments, including hydrodynamic, thermohydraulic and technological stands, has been created at SSC RF - IPPE. In addition, JSC “SSC RF - IPPE” has a complex of fast physical stands, including two critical stands - BFS-1 and the world's largest physical stand BFS-2. The article presents the characteristics and the possibility of stands designed for research in the field of hydrodynamics, heat transfer and coolant technology in support of design solutions, safety improvement and testing of equipment elements and assemblies of operating and planned installations with fast reactors with sodium, lead and lead-bismuth coolants, as well as for accelerator-controlled systems and thermonuclear fusion, low-power nuclear power plants for space: - Hydrodynamic stands - “SGDI” (air), “V-2” (air), “SGI” (water), “V-200” (water), “GDK” (air). - Thermal-hydraulic liquid metal stands - “6B” (Na, Na-K), “AR-1” (Na, Na-K), “Pluton” (Na), “SPRUT” (Na, Na-K, Pb, Pb-Bi, water). - Technological liquid metal stands - “Protva-1” (Na), “Protva-2” (Na), “PUSHM” (Na), “Armatura” (Na), “IK-MT” (Na), “SID” (Na), “BTS” (Na), “TT-1M” (Pb), “TT-2M” (Pb-Bi), “LIS-M” (Li). A large-scale sodium test stands “SAZ” is under construction, which allows testing full-scale prototypes of equipment and its elements to substantiate existing and future projects of fast sodium reactors. The BFS complex of physical stands is the world's only experimental tool for full-scale modeling of the cores of nuclear reactors of various types (of any composition, geometry and configuration). The materials and construction of the stands allow simulating the core, breeding zones, reflectors and in-core shielding, as well as elements of fuel cycles and storage facilities for spent nuclear fuel and radioactive waste. Reactor materials of the stands (metallic plutonium, oxide and metallic highly enriched uranium with enrichment of 36% and 90% in uranium-235, hundreds of tons of fertile materials, construction materials, various coolants) make it possible to assemble both complex full-scale models of fast reactors, and benchmarks, experiments for which are carried out to correct neutron-physical constants and improve computational methods.

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40

Okunev, Viacheslav. "The concept of a fast reactor with liquid metal fuel in tungsten capsules". E3S Web of Conferences 411 (2023): 01013. http://dx.doi.org/10.1051/e3sconf/202341101013.

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The concept of a dual-purpose high-power nuclear reactor is proposed. One of the goals is the production of electricity, the other is the production of high-potential thermal energy. It is proposed to use liquid fuel based on waste uranium and plutonium extracted from the spent fuel of VVER reactors (purified from the 238Pu isotope). The fuel is in sealed tungsten capsules. Lead extracted from thorium ores is used to cool the reactor. The electrical power of the reactor is 3.3 GW. The layout of the reactor is identical to the BREST-OD-300 reactor under construction. The analysis of emergency modes from among ATWS (anticipated transients without scram) is carried out. The reactor is reliable and safe. The maximum temperature of a high-temperature reactor coolant is close to the boiling point of lead. By the nature of the change in the maximum temperatures of the core components, the reactor occupies an intermediate position between a reactor with solid metallic fuel and a reactor with cermets based on UN-PuN and metal uranium nanopowder.

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41

Ben-Dor, G., A. Dubinsky y T. Elperin. "Optimization of multi-layered metallic shield". Nuclear Engineering and Design 241, n.º 6 (junio de 2011): 2020–25. http://dx.doi.org/10.1016/j.nucengdes.2011.01.046.

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42

Pasquato, Leone, Christoph Strangfeld, Esko Strömmer, Sergej Johann, Vera Lay, Christian Köpp y Ernst Niederleithinger. "Embedded sensors system to monitor cemented waste drums". Safety of Nuclear Waste Disposal 2 (6 de septiembre de 2023): 21. http://dx.doi.org/10.5194/sand-2-21-2023.

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Abstract. BAM (Federal Institute for Materials Research and Testing) is developing an electronic measurement system to be placed inside a waste drum, which will be filled with concrete. The goal of this measurement system is to monitor the process of hardening and the evolution of the concrete itself over time to indirectly identify potential defects such as corrosion or cracking. The measured parameters are humidity, temperature, and pressure. In this regard, particular attention was given to the design of the electronic board's enclosure, to allow the sensors to measure the state of the concrete without being in direct contact with it. In the scope of the European Commission's project of PREDIS, the supply of power to the battery-less sensors and the data acquired by such sensors are transmitted through the metallic waste drum by an innovative wireless technology developed by VTT (Technical Research Centre of Finland Ltd) in order to ensure long-term operation, while keeping the integrity of the sealed container. The sensing system is made of a chain of small units, called SensorNodes. Each SensorNode includes two off-the-shelf sensors, with one for relative humidity and temperature and one for pressure and temperature. A SensorNode is designed to have a unique identifier, in order to be connected to other units while being uniquely discoverable by a standard communication protocol. In this way, a distributed matrix of measurement points is created. One of the most challenging tasks in designing a measurement system to run in a harsh environment (such as hardening concrete) is to let the sensors sense the external environment without damaging the sensors themselves. In order to keep the external environment away from the electronic board while still letting the sensors measure the concrete behaviour, holes have been drilled through the lid and covered from the inside with a layer of porous membrane. The membrane's pores allow water and gas particles to pass through and let the enclosed air equilibrate with the external environment. With the help of the developed sensors, monitoring concrete in cemented waste drums will be possible. The derived data will also serve as the basis for ongoing modelling approaches for digital twins within the PREDIS project. Overall, the sensors provide a means of enabling safe nuclear waste management through advanced monitoring.

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43

Bieniussa, K. W. "German codes and standards concerning metallic nuclear power plant components — Present state and trends expected". Nuclear Engineering and Design 98, n.º 3 (enero de 1987): 279–81. http://dx.doi.org/10.1016/0029-5493(87)90004-5.

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Tachibana, Yukio y Tatsuo Iyoku. "Structural design of high temperature metallic components". Nuclear Engineering and Design 233, n.º 1-3 (octubre de 2004): 261–72. http://dx.doi.org/10.1016/j.nucengdes.2004.08.013.

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45

Sulatsky, Andrey A., Viacheslav I. Almjashev, Vladimir S. Granovsky, Vladimir B. Khabensky, Evgeniy V. Krushinov, Sergey A. Vitol, Victor V. Gusarov et al. "Experimental study of oxidic-metallic melt oxidation". Nuclear Engineering and Design 363 (julio de 2020): 110618. http://dx.doi.org/10.1016/j.nucengdes.2020.110618.

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46

Rothfuss, Helmut, Dieter Stausebach y Manfred Ullrich. "Metallic core internals of the modular HTR". Nuclear Engineering and Design 147, n.º 1 (enero de 1994): 93–100. http://dx.doi.org/10.1016/0029-5493(94)90260-7.

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47

Xie, Yi, Jinsuo Zhang, Xiang Li, Jeremy P. Isler, Michael T. Benson, Robert D. Mariani y Cetin Unal. "Lanthanide migration and immobilization in metallic fuels". Progress in Nuclear Energy 109 (noviembre de 2018): 233–38. http://dx.doi.org/10.1016/j.pnucene.2018.08.019.

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Ebert, Elena L., Andrey Bukaemskiy, Fabian Sadowski, Steve Lange, Andreas Wilden y Giuseppe Modolo. "Reprocessability of molybdenum and magnesia based inert matrix fuels". Nukleonika 60, n.º 4 (1 de diciembre de 2015): 871–78. http://dx.doi.org/10.1515/nuka-2015-0124.

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Abstract This work focuses on the reprocessability of metallic 92Mo and ceramic MgO, which is under investigation for (Pu,MA)-oxide (MA = minor actinide) fuel within a metallic 92Mo matrix (CERMET) and a ceramic MgO matrix (CERCER). Magnesium oxide and molybdenum reference samples have been fabricated by powder metallurgy. The dissolution of the matrices was studied as a function of HNO3 concentration (1-7 mol/L) and temperature (25-90°C). The rate of dissolution of magnesium oxide and metallic molybdenum increased with temperature. While the MgO rate was independent of the acid concentration (1-7 mol/L), the rate of dissolution of Mo increased with acid concentration. However, the dissolution of Mo at high temperatures and nitric acid concentrations was accompanied by precipitation of MoO3. The extraction of uranium, americium, and europium in the presence of macro amounts of Mo and Mg was studied by three different extraction agents: tri-n-butylphosphate (TBP), N,Nʹ-dimethyl-N,Nʹ-dioctylhexylethoxymalonamide (DMDOHEMA), and N,N,N’,N’- -tetraoctyldiglycolamide (TODGA). With TBP no extraction of Mo and Mg occurred. Both matrix materials are partly extracted by DMDOHEMA. Magnesium is not extracted by TODGA (D < 0.1), but a weak extraction of Mo is observed at low Mo concentration.

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., Manisha, Anima Sunil Dadhich y Anik Sen. "Preparation of barium ferrate from mill scale and degradation study of aqueous Eriochrome black – T". Research Journal of Chemistry and Environment 27, n.º 2 (15 de enero de 2023): 10–19. http://dx.doi.org/10.25303/2702rjce10019.

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This present study involves the utilization of mill scale, an iron rich steel industry waste, to synthesize barium ferrate. Ferrate with iron in +(VI) oxidation state is a green chemical with promising oxidation, disinfection and coagulation properties. The X-Ray diffraction(XRD) analysis of mill scale indicates the presence of iron in three forms hematite, magnetite and wustite. Mill scale was dissolved in concentrated HCl. The metallic composition of dissolved mill scale was determined by Inductively Couple Plasma Mass Spectroscopy(ICP-MS). Barium ferrate was synthesized using the dissolved mill scale by the wet oxidation method. The resulting crude ferrate was washed with different organic solvents to remove the impurities. The product thus obtained was analyzed by chromite method and characterized by Fourier Transform Infrared Spectroscopy(FTIR). Degradation of aqueous Eriochrome Black – T(EBT) with barium ferrate was studied at different pH, time and concentration. The degradation of the dye was monitored by UV-Visible spectroscopy and chemical oxygen demand. The degradation products of the aqueous EBT were characterized by Gas Chromatography-Mass Spectroscopy(GC-MS) and Proton Nuclear Magnetic Resonance (1H-NMR). Computational studies were performed to compare the stability of barium ferrate with that of potassium ferrate. The 1H-NMR calculations were performed to compare the degradation products obtained experimentally.

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Yacout, Abdellatif M., Kun Mo, Aaron Oaks, Yinbin Miao, Tanju Sofu y Walid Mohamed. "FIPD: The SFR metallic fuels irradiation & physics database". Nuclear Engineering and Design 380 (agosto de 2021): 111225. http://dx.doi.org/10.1016/j.nucengdes.2021.111225.

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