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Artykuły w czasopismach na temat "Metallic nuclear waste"
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Stoulil, J., i D. Dobrev. "Microbial corrosion of metallic materials in a deep nuclear-waste repository". Koroze a ochrana materialu 60, nr 2 (1.06.2016): 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., i D. D. Keiser. "TEM Analysis of Corrosion Products From a Radioactive Stainless Steel-based Alloy". Microscopy and Microanalysis 6, S2 (sierpień 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., i D. D. Keiser. "Actinides in metallic waste from electrometallurgical treatment of spent nuclear fuel". JOM 55, nr 9 (wrzesień 2003): 59–60. http://dx.doi.org/10.1007/s11837-003-0032-z.
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Holt, Erika, Maria Oksa, Matti Nieminen, Abdesselam Abdelouas, Anthony Banford, Maxime Fournier, Thierry Mennecart i 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 i Clint A. Sharrad. "A review of contamination of metallic surfaces within aqueous nuclear waste streams". Progress in Nuclear Energy 159 (maj 2023): 104637. http://dx.doi.org/10.1016/j.pnucene.2023.104637.
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Moiseenko, V., i S. Chernitskiy. "Nuclear Fuel Cycle with Minimized Waste". Nuclear and Radiation Safety, nr 1(81) (12.03.2019): 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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Janney, Dawn E. "Host phases for actinides in simulated metallic waste forms". Journal of Nuclear Materials 323, nr 1 (listopad 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, nr 3 (1.02.2014): 503–25. http://dx.doi.org/10.1007/s11837-014-0873-7.
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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 (sierpień 2002): 1310–11. http://dx.doi.org/10.1017/s1431927602104983.
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Pavliuk, Alexander O., Evgeniy V. Bespala, Sergey G. Kotlyarevskiy, Ivan Yu Novoselov i Veleriy N. Kotov. "Analysis of Heat Release Processes inside Storage Facilities Containing Irradiated Nuclear Graphite". Science and Technology of Nuclear Installations 2022 (30.01.2022): 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.
Rozprawy doktorskie na temat "Metallic nuclear waste"
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Fager, Fredrick, i Serg Chanouian. "Nuclear Waste Canister : Evaluating the mechanical properties of cassette steel after casting". Thesis, KTH, Materialvetenskap, 2017. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-209803.
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Företaget Svensk Kärnbränslehantering AB (SKB) håller på att utveckla en slutförvaringskapsel som kommer innehålla avfall från den svenska kärnkraften. Det är dock fortfarande en process under utveckling och därför undersöks olika typer av metoder och kapselmaterial för att kunna tillverka en hållbar och säker kapsel. Kapseln består av ett hölje av kopparrör med svetsad botten och lock och en insats med stållock. Insatsen är en cylindrisk konstruktion av segjärn och innehåller en svetsad stålkassett för att skapa utrymmen till det använda kärnbränslet. Insatsen innehåller bland annat stålrör som under tillverkning får utstå en gjutprocess med segjärn och erhåller efter det icke homogena egenskaper. Målet med undersökningen är hur stor påverkan gjutningen har på stålets kemiska sammansättning samt mikrostrukturer. Det som orsakar de inhomogena egenskaperna är främst värmebehandlingen som driver diffusionen av kol från gjutjärnet till stålet, som då ger ett hårdare men sprödare material. Med hjälp av experiment och simuleringar upptäcks hur mycket kol som diffunderar in i stålet samt ändringar i den kemiska sammansättningen i de påverkade zonerna. Identifiering av fasomvandlingar, diffusion och ändringar i mikrostrukturer är stora faktorer som i sin tur ändrar de mekaniska egenskaperna i stålet.The Swedish Nuclear Fuel and Waste Management Company (SKB) have developed a final storage canister that will contain waste from the Swedish nuclear power plants. However, it is still in a development phase and therefore different types of methods and canister materials are investigated to produce the most durable and safe canister. The canister is made of a copper tube with a welded bottom and lid with an insert. The insert is a cylindrical construction of nodular cast iron that contains a welded steel cassette, to make space for the spent fuel, and a steel lid. The steel tubes showed inhomogeneous properties after being exposed to a casting around them. The aim of this investigation is to clarify the impact of casting on the chemical composition of the steel as well as the microstructure. The cause to the inhomogeneous properties were the diffusion of carbon from the cast iron to the steel, which then produced a harder and more brittle material. Experiments and simulations were used to see the carbon diffusion into the steel as well as what happens with the chemical composition in the affected zones. Identification of phase changes, diffusion and microstructures contributed to changes of mechanical properties in the steel.
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Hamza, Letissia. "Réactivité du graphite, magnésium et uranium, déchets nucléaires des réacteurs UNGG, dans des hydroxydes fondus". Electronic Thesis or Diss., université Paris-Saclay, 2025. http://www.theses.fr/2025UPASP004.
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Cette thèse s’inscrit dans le cadre du projet SELF France 2030, qui propose un prétraitement oxydant par l’eau des déchets métalliques nucléaires dans un sel NaOH-KOH-H₂O fondu à 225⁰C, avant leur conditionnement en matrices cimentaires. Ce procédé permet d’inerter les déchets métalliques et assure la sûreté du colis de stockage. Le travail de thèse a donc été consacré à l’étude de la réactivité des principaux déchets nucléaires métalliques – graphite, magnésium et uranium – dans le sel fondu NaOH-KOH à 225⁰C. Une première partie concerne le mélange NaOH-KOH (51,5 – 48,5 mol%) fondu à 225⁰C pour acquérir des données thermodynamiques et expérimentales sur sa stabilité chimique et électrochimique. Les études gravimétriques et électrochimiques montrent que le mélange contient 15 mol% d’eau à 225⁰C, ce qui est caractéristique d’une forte solvatation de l’eau dans ce milieu. Les propriétés redox du sel ont été étudiées par électrochimie couplée à la micro-chromatographie en phase gazeuse (μGC). Les réactions aux limites anodique et cathodique sont fortement influencées par la teneur en eau : en milieu hydraté, la limite cathodique correspond à la réduction de H₂O en H₂, alors qu'en milieu déshydraté, c’est la réduction de Na⁺ qui limite le domaine d’électroactivité. Quant à la limite anodique, elle est attribuée à l’oxydation de OH⁻ en O₂ dans les milieux contenant des teneurs en eau supérieures à 8,7 mol% et à l’oxydation de OH⁻ en O₂⁻ pour des teneurs inférieures. L’étude électrochimique a permis de calculer le coefficient d’activité de l’eau dans le mélange d’hydroxydes fondus à 225⁰C. Le résultat confirme une forte solvatation de l’eau dans ce milieu. L’eau étant l’élément oxydant dans le milieu, il était important de pouvoir suivre sa teneur in-situ. C’est ainsi qu’une droite de calibration basée sur la mesure du courant de pic de réduction de l’eau - Ipc =f([H₂O]) - a été établie pour pouvoir doser l’eau lors des différentes expériences. Le coefficient d’activité de NaOH a également été déterminé par électrochimie. Cet ensemble de données expérimentales a permis de calculer le diagramme de stabilité du sel NaOH-KOH en fonction du potentiel et de la teneur en eau. Enfin, un moyen de contrôle et de maintien de la quantité d’eau a été proposé pour assurer une oxydation continue des déchets métalliques. La deuxième partie de cette thèse a été dédiée à l'étude de la réactivité du graphite, du magnésium et de l’uranium dans les hydroxydes fondus à 225⁰C contenant de l’eau. Cette étude a montré que le graphite est stable dans les hydroxydes fondus. D’après les données thermodynamiques, le magnésium peut être oxydé par l’eau ou les ions Na⁺, conduisant respectivement à la formation de H₂ ou de Na, composé pyrophorique. Cependant, les études expérimentales montrent que le magnésium est toujours oxydé par H₂O et, par électrochimie, on montre que l’oxydation du magnésium est observée à un potentiel supérieur à celui de la réduction des ions Na⁺. Par ailleurs, la cinétique d’oxydation de Mg est proportionnelle à la concentration en eau pour les plus faibles teneurs en eau, ce qui est caractéristique d’un contrôle cathodique de la dissolution. A des teneurs élevées en eau, la cinétique d’oxydation atteint une limite qui dépend de la surface active du magnésium, on a alors un contrôle anodique de la dissolution. Pour la mise en œuvre industrielle et afin de gérer la cinétique de dissolution des déchets et la formation de H₂, il est donc préconisé de partir d’un sel fondu déshydraté et d’ajouter l’eau en continu. L’ensemble des données expérimentales a permis de calculer le diagramme de stabilité du magnésium dans le milieu. Concernant le comportement de l’uranium dans le sel fondu, les calculs thermodynamiques montrent que l'uranium est oxydé préférentiellement en K₂UO₄. Les analyses par chromatographie gazeuse ont confirmé l’oxydation de l’uranium par l’eau et l’étude expérimentale a permis de proposer un mécanisme réactionnelThis thesis is part of the SELF France 2030 project, which proposes a water-based oxidative pretreatment of metallic nuclear waste in a NaOH-KOH-H₂O salt melted at 225⁰C before conditioning in cementitious matrices. This process inserts the metal waste and ensures the safety of the disposal package. The thesis work was therefore devoted to studying the reactivity of the main metallic nuclear wastes - graphite, magnesium, and uranium - in molten NaOH-KOH salt at 225⁰C. The first part concerns the NaOH-KOH mixture (51.5 - 48.5 mol%) melted at 225⁰C to acquire thermodynamic and experimental data on its chemical and electrochemical stability. Gravimetric and electrochemical studies show that the mixture contains 15 mol% water at 225⁰C, which is characteristic of high water solvation in this medium. The redox properties of the salt were studied by electrochemistry coupled with gas-phase micro-chromatography (μGC). Reactions at the anodic and cathodic limits are strongly influenced by water content: in hydrated media, the cathodic limit corresponds to the reduction of H₂O to H₂, whereas in dehydrated media, it is the reduction of Na⁺ that limits the electroactivity range. The anodic limit is attributed to the oxidation of OH⁻ to O₂ in media with water contents above 8.7 mol% and to the oxidation of OH- to O₂⁻ for lower contents. The electrochemical study enabled us to calculate the water activity coefficient in the molten hydroxide mixture at 225⁰C. The result confirms the high solvation of water in this medium. Water is the oxidizing element in the medium, so it was important to monitor its content in situ. Therefore, a calibration line based on measuring the water reduction peak current - Ipc =f([H₂O]) - was established, enabling water to be measured in the various experiments. The activity coefficient of NaOH was also determined electrochemically. This set of experimental data was used to calculate the stability diagram of the NaOH-KOH salt as a function of potential and water content. Finally, controlling and maintaining the amount of water was proposed to ensure continuous oxidation of metallic waste. The second part of this thesis was dedicated to studying the reactivity of graphite, magnesium, and uranium in molten hydroxides containing water at 225⁰C. This study showed that graphite is stable in molten hydroxides. This study showed that graphite is stable in molten hydroxides. According to thermodynamic data, magnesium can be oxidized by water or Na⁺ ions, forming a pyrophoric compound of H₂ or Na. However, experimental studies show that H₂O continuously oxidizes magnesium, and electrochemically, magnesium oxidation is observed at a higher potential than Na⁺ ion reduction. Furthermore, Mg oxidation kinetics are proportional to water concentration at lower water contents, characteristic of cathodic dissolution control. At higher water contents, the oxidation kinetics reach a limit that depends on the active surface of the magnesium, giving anodic control of the dissolution. For industrial implementation and to manage waste dissolution kinetics and H₂ formation, it is recommended to start with a dehydrated molten salt and continuously add water. We calculated a stability diagram for magnesium in the medium based on all these experimental data. Concerning the behavior of uranium in molten salt, thermodynamic calculations show that uranium is preferentially oxidized to K₂UO₄. Gas chromatography analyses confirmed uranium oxidation by water, and experimental studies proposed a reaction mechanism
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Ngwenya, Nonhlanhla. "Bioremediation of metallic fission products in nuclear waste : biosorption and biorecovery". Thesis, 2011. http://hdl.handle.net/2263/28663.
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The performance of a growing sulphate reducing bacteria consortium for Sr2+, Co2+ and Cs+ removal from solution in a batch sulphidogenic bioreactor was investigated. Metal removal by the growing bacterial consortium, and microbial culture growth and metabolic activities (biological sulphate removal) were continuously monitored in the bioreactors over the duration of the treatment period. On the other hand, diversity changes within the bacterial consortium before and after bioreactor operation (28 days) were performed using the partial 16S rRNA fingerprinting method. In the original bacterial consortium, Enterococcus and Staphylococcus sp. were the dominant bacterial species. However, the presence of Sr2+, Co2+ and Cs+ in the growth media, resulted in the emergence of new bacterial species belonging to the Citrobacter, Paenibacillus, and Enterococcus and Stenotrophomonas genera, respectively. The Citrobacter and Paenibacillus sp. demonstrated high tolerance towards the presence of the divalent cations, Sr2+ and Co2+, respectively, while the Enterococcus and Stenotrophomonas sp., demonstrated Cs+ high tolerance. The bacterial growth and sulphate removal rate were significantly decreased at initial metal ion concentrations ≥100 mg/L. The toxicity and inhibitory effects of the metals on the present SRB consortium was observed in the order Sr>Co>Cs. The metal uptake capacity (qτ) of the bacterial consortium decreased with increasing initial metal concentration, and complete Sr2+, Co2+ and Cs+ removal was observed at initial metal concentrations ≤75 mg/L. Overall, the present SRB consortium demonstrated a superior Sr2+ removal capacity (qmax= 405 mg/g), and the least for Cs2+, where qmax = 192 mg/g. The present SRB culture exhibited a superior Sr+ and Cs+ binding capacity, compared to other studies in literature. Results from Sr2+, Co2+ and Cs+ biosorption kinetics indicate that initial concentration and solution pH played a vital role in determining the rate of metal removal kinetics. The experimental data was successfully analysed by the pseudo-second-order rate model, demonstrating that chemisorption is the main rate limiting step for the removal of Sr2+, Co2+ and Cs+ from solution. In this study, the adsorption behaviour of protons and of Sr2+, Co2+ and Cs+ onto the bacterial consortium cell surfaces was evaluated under anaerobic conditions as a function of pH (4-10), ionic strength (0.01, 0.05, 0.1M) and temperature (25, 50 and 75°C). Acid-base titrations of the bacterial suspension indicated that the titration data could be adequately described by a four site non-electrostatic model, with pKa values of 4.41, 6.69, 8.10 and 10. The Sr2+, Co2+ and Cs+ adsorption data could be fitted with a two site non-electrostatic model, involving the type 1 and 2 sites (carboxylic and phosphoryl sites). Increasing the ionic strength had a negative effect on the adsorption of metal ions from solution. There was no observed temperature dependence on the adsorption of Co2+ and Cs+ from solution. In summary, results obtained in this study have shown that the processes involved in microbial Sr2+, Co2+ and Cs+ removal from contaminated sources is a direct function of the microbial characteristics and efficiency, mass transfer and surface complexation effects under varying environmental conditions. One important goal to be achieved in future studies will be the determination of the intrinsic stability constants and the structure of the formed metal complexes species. These constants can be used directly in risk assessment programs.Thesis (PhD(Eng))--University of Pretoria, 2011.
Chemical Engineering
unrestricted
Książki na temat "Metallic nuclear waste"
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Canada, Atomic Energy of. Metallic Iron Content of Candidate Clays and Silica Sand For Use in the Canadian Nuclear Fuel Waste Management Program. S.l: s.n, 1985.
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Pflugrad, K., i D. Hofman. Technical Seminar on Melting and Recycling of Metallic Waste Materials from Decommissioning of Nuclear Installations: 26 to 29 October 1993. European Communities / Union (EUR-OP/OOPEC/OPOCE), 1994.
Znajdź pełny tekst źródłaCzęści książek na temat "Metallic nuclear waste"
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Ishii, Kyoko, Mitsuaki Yamaoka, Yasuyuki Moriki, Takashi Oomori, Yasushi Tsuboi, Kazuo Arie i Masatoshi Kawashima. "Development of Uranium-Free TRU Metallic Fuel Fast Reactor Core". W Nuclear Back-end and Transmutation Technology for Waste Disposal, 155–67. Tokyo: Springer Japan, 2014. http://dx.doi.org/10.1007/978-4-431-55111-9_15.
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Ebert, W. L. "Metallic Waste Forms". W Comprehensive Nuclear Materials, 505–38. Elsevier, 2012. http://dx.doi.org/10.1016/b978-0-08-056033-5.00109-9.
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Reza Doust, M. "Metallic Nanoparticles in the Glasses: Advances and Current Challenges". W Materials Research Foundations, 257–78. Materials Research Forum LLC, 2024. http://dx.doi.org/10.21741/9781644903056-7.
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Glasses are fascinating materials with diverse applications. Rare earth doped glasses are well-known for their optical properties which could be used as solid-state lasers, optical sensors, scintillators, optical thermometers, optical fibers in telecommunications, nuclear waste storages, etc. Recently, by increasing the interest in the modification of materials on the nanoscale, glasses doped with metallic nanoparticles have attracted much attention. This type of doping was used historically in the coloration of the glasses, however, in the new millennium it is used also as the optical centers to enhance the radiative quantum yield of the luminescent materials. In this proposal, yet there have been confronted several challenges which need further investigations. For example, not all the cases of metal particle doping in luminescence materials yield an enhancement in emission intensities of lanthanide ions in glasses. This chapter revisits the important experimental results and discusses them as three different arguments.
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Féron, Damien. "Lifetime prediction of metallic barriers in nuclear waste disposal systems: overview and open issues related to sulphur-assisted corrosion". W Sulphur-assisted corrosion in nuclear disposal systems, 66–80. CRC Press, 2020. http://dx.doi.org/10.1201/9781003059448-4.
Pełny tekst źródłaStreszczenia konferencji na temat "Metallic nuclear waste"
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Coltella, Thomas, Francesca Valente, Veronica Pierantoni, Cristina Ricci, Michele Frignani, Monica Linda Frogheri, Matteo Di Prinzio, Mario Mariani, Elena Macerata i Simone Tiozzo. "Nuclear Waste Treatment: Vitrification of Iron-Phosphate Sludge". W ASME 2023 International Conference on Environmental Remediation and Radioactive Waste Management. American Society of Mechanical Engineers, 2023. http://dx.doi.org/10.1115/icem2023-110227.
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Abstract Within the framework of developing advanced processes for nuclear waste treatment, Ansaldo Nucleare is developing and testing an innovative technology to perform the conditioning of radioactive ferrous waste materials, classified as Low and Intermediate Level Waste (LILW), resulting from decommissioning activities of Nuclear Facilities. This technology is based on a patented process dedicated to the treatment of decontamination solutions coming from the pickling process of the metallic components, with the aim to minimize the waste to be disposed, in an inert final product, and maximize the freely released material, that once it is decontaminated can be recycled. The process results in the production of a radioactive sludge, made mostly of iron-phosphate salts, which can be thermally treated to produce a ready-to-storage, homogenous and chemically resistant glass product. The process was first tested at laboratory scale and then in a pilot plant, installed at Politecnico di Milano. The main strengths of this innovative technology are: • Lower amount of fresh materials/chemicals requested for the waste treatment, resulting in lower costs associated with the process, also thanks to the continuous recycling of the pickling solution; • Lower volume and higher long-term stability of the nuclear waste for the final storage; • Facility with reduced footprint and transportable, designed to be installed in commercial containers.
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Moggia, Fabrice, i Xavier Lecardonnel. "Metallic Surfaces Decontamination by Using LASER Light". W ASME 2013 15th International Conference on Environmental Remediation and Radioactive Waste Management. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icem2013-96301.
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Metal surface cleaning appears to be one of the major priorities for industries especially for nuclear industries. The research and the development of a new technology that is able to meet the actual requirements (i.e. waste volume minimization, liquid effluents and chemicals free process…) seems to be the main commitment. Currently, a wide panel of technologies already exists (e.g. blasting, disk sander, electrodecontamination…) but for some of them, the efficiency is limited (e.g, Dry Ice blasting) and for others, the wastes production (liquid and/or solid) remains an important issue. One answer could be the use of a LASER light process. Since a couple of years, the Clean-Up Business Unit of the AREVA group investigates this decontamination technology. Many tests have been already performed in inactive (i.e. on simulants such as paints, inks, resins, metallic oxides) or active conditions (i.e. pieces covered with a thick metallic oxide layer and metallic pieces covered with grease). The paper will describe the results obtained in term of decontamination efficiency during all our validation process. Metallographic characterizations (i.e. SEM, X-ray scattering) and radiological analysis will be provided. We will also focus our paper on the future deployment of the LASER technology and its commercial use at La Hague reprocessing facility in 2013.
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Rodri´guez, M. "Recycling of Metallic Waste Produced During the Decommissioning of Vandellos 1 NPP". W ASME 2003 9th International Conference on Radioactive Waste Management and Environmental Remediation. ASMEDC, 2003. http://dx.doi.org/10.1115/icem2003-4942.
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The decommissioning of a Nuclear Power Plant leads to the generation of a high volume of metallic waste, of which a large quantity can be recycled, after separating it from radioactive metal. The methodology that allows to separate radioactive metal from non-radioactive metal is an essential part of Decommissioning, and it requires complex equipment, procedures and controls, both inside Enresa and external (Regulatory Body). After undergoing these controls, most of the metallic material is sent to recycling facilities, where it is mixed with a much larger proportion of metal coming from conventional scrap yards, both here in Spain and abroad. In recent years, there have been a few incidents in melting plants, due to the presence of undetected radioactive material among certain batches of scrap metal. In order to tackle the public concern associated with potential risk, a series of measures have been designed to prevent these incidents or minimise their effects, should they occur. The following presentation will first describe the Spanish protocol established by different national institutions to prevent the presence of radioactive waste among the raw material for recycling, should this occur detect and control it, and secondly the methodology that guarantees that Enresa does not erroneously send radioactive material arising from the decommissioning of nuclear power plants to any smelting facility.
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Branagan, D., J. Buffa i M. Maston. "Advanced Nanoscale Neutron Absorber Coatings for Safe Longterm Storage of Spent Nuclear Fuel and Nuclear Waste". W ITSC2005, redaktor E. Lugscheider. Verlag für Schweißen und verwandte Verfahren DVS-Verlag GmbH, 2005. http://dx.doi.org/10.31399/asm.cp.itsc2005p0551.
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Abstract Neutron absorbers are expected to play an important role in the long-term storage of spent nuclear fuels and nuclear wastes. High neutron absorbing capability, long-term stability, and the capacity to stay with the fuel are important criteria in preventing critical conditions during possible waste package degradation in geological time frames. Existing available neutron absorbing materials are based on boron or boron-10 isotope modifications of austenitic stainless steels or to aluminum based metal matrix composites. Specific rare earths such as gadolinium, samarium, or europium are found to have much higher thermal neutron cross section than boron or boron-10 but have high reactivity which limit their stability and ultimate applicability. In this paper, it is described how it is possible through a nanotechnology approach, to overcome the solubility and stability limitations of conventional materials to allow incorporation of high amounts of boron and rare earths into advanced HVOF coatings. During the development of the NeutraShieldTM Coatings, it was found that high fractions of rare earth elements such as gadolinium along with high concentrations of boron could be dissolved in the liquid melt and then remain soluble in the metallic glass structure. During the transformation of the glass to the nanocomposite structure, the rare earths are found to come out of supersaturated solid solution to form stable nanoscale ternary intermetallic R2Fe14B phases which form in a commensurate fashion and is protected by the highly noble matrix. Abstract only; no full-text paper available.
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James, Randy J., Kenneth Jaquay i Michael J. Anderson. "Design by Analysis of Waste Packages at Yucca Mountain for Impact Loads". W 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75355.
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The proposed geologic repository under development at Yucca Mountain, Nevada, will employ multiple shell metallic containers (waste packages) for the disposal of nuclear waste. The waste packages represent a primary engineered barrier for protection and containment of the radioactive waste, and the design of these containers must consider a variety of structural conditions to insure structural integrity. Some of the more challenging conditions for structural integrity involve severe impact loading due to hypothesized event sequences, such as drops or collisions during transport and placement. Due to interactions between the various components leading to complex structural response during an impact sequence, nonlinear explicit dynamic simulations and highly refined models are employed to qualify the design for these severe impact loads. This paper summarizes the Design by Analysis methodologies employed for qualification of waste package design under impact loading and provides several illustrative examples using these methods. Example evaluations include a collision of a waste package by the Transport and Emplacement Vehicle (TEV) and two scenarios due to seismic events, including WP impact within the TEV and impact by falling rock. The examples are intended to illustrate the stringent Design by Analysis methods employed and also highlight the scope of structural conditions included in the design basis for waste packages to be used for proposed nuclear waste storage at Yucca Mountain.
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Slimak, Andrej, i Vladimir Necas. "The Analysis of Metal Melting Application in the Management of Metallic Radioactive Materials Arising From Decommissioning of Nuclear Installations". W ASME 2013 15th International Conference on Environmental Remediation and Radioactive Waste Management. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icem2013-96143.
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Basic characterization of the waste management process during decommissioning of nuclear installations is described in the presented paper. A brief description is given of conditional and unconditional release of materials into the environment. The paper deals also with metal melting as prospective decontamination technique which can significantly reduce metallic radioactive waste. The material and radioactivity flow in the decommissioning process should be followed using the integrated material flow tool that is implemented into the standardized analytical decommissioning parameters calculation code OMEGA. Applying the integrated material flow tool, it is possible to monitor radiological and physical properties of individual material items listed in the nuclear installation input database, from dismantling up to their release into the environment or disposal in repository. Using the OMEGA code and two model databases, several scenarios related to metal melting are evaluated. The impact of applying different input decommissioning parameters on the metal distribution is the main result discussed in the paper.
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Dutzer, Michel, Ge´rald Ouzounian, Roberto Miguez i Jean-Louis Tison. "Radioactive Waste: Feedback of 40-Year Operations in France". W ASME 2010 13th International Conference on Environmental Remediation and Radioactive Waste Management. ASMEDC, 2010. http://dx.doi.org/10.1115/icem2010-40081.
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France’s experience in the management of radioactive waste is supported by forty years of operational activities in the field of surface disposal. This feedback is related to three disposal facilities: Centre de la Manche disposal, not far away Cherbourg, from design to post-closure facility. Centre at Soulaines-Dhuys from site selection to design to operation during nearly 20 years. Centre at Morvilliers from site selection to operation for seven years now. During the operational period of Centre de la Manche disposal facility (1969–1994), the safety concept for low- and intermediate level short lived waste (LIL-SLW) was developed and progressively incorporated in the procedures of the facility. The facility entered its institutional control period and the experience of this facility has been useful for the operating facilities. Centre de l’Aube that took over Centre de la Manche, and Morvilliers for very low level wastes. Both facilities currently accommodate the major part of the volume of radioactive wastes that are generated in France. However disposal facilities have to be considered as rare resources. Then new waste management options are being investigated as the disposal of large components or recycling metallic wastes within the nuclear industry.
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Cournoyer, Michael E., Timothy P. Martinez i Robert F. Grundemann. "Waste Avoidance Program for Mercury-Laden Mixed Waste". W ASME 2001 8th International Conference on Radioactive Waste Management and Environmental Remediation. American Society of Mechanical Engineers, 2001. http://dx.doi.org/10.1115/icem2001-1300.
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Abstract Mercury is ideal for use in such things as thermometers and vacuum gauges (manometers) because this naturally occurring silvery liquid expands and contracts evenly with temperature and pressure changes. Within the Nuclear Materials Technology (NMT) Division of Los Alamos National Laboratory, mercury containing devices are used for a variety of operations, including actinide chemistry, weapons production, radiochemistry, and analytical chemistry. Mercury present in these instruments does not in itself constitute a risk of contamination since the metal is contained within a closed system. However, breakage, inadequate maintenance and disposal of such instruments can expose workers and the public to this toxic substance. In this study, waste minimization issues associated with replacing these mercury-laden instruments with non- and less-hazardous devices are addressed. These include institutional program available to support this effort, the hazards grouped with mercury, and devices that use mercury. Life cycle management issues are also examined. Procedures and waste minimization processes used to remove the metallic mercury from an operation will be presented. A project in which mercury containing thermometers and manometers are systematically replace with less hazardous instruments is discussed. This project resulted in approximately 5 kilograms of mercury being removed from Radiological Controlled Areas during project, which represents the elimination of a potential liability of the generation of 100 m3 of Mixed Low-Level Waste in the future. As a final step, an approach to get buy-in from both technical and budget minded employees is presented. The elimination and substitution of hazardous materials in NMT Division is an ongoing process that starts during the design phase of a process or facility and continues through the performance of routine procedures.
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Jaquay, Kenneth R., i Michael J. Anderson. "Yucca Mountain Project Structural Fragility Estimates for Impact Loading of Waste Packages". W ASME 2008 International Mechanical Engineering Congress and Exposition. ASMEDC, 2008. http://dx.doi.org/10.1115/imece2008-66538.
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A methodology is presented for estimating the ultimate structural capability (fragility) of metallic nuclear waste disposal containers (waste packages) subject to impact events. The LS-DYNA finite element analysis (FEA) computer code and massively parallel processing (MPP) is used for nonlinear, dynamic-plastic, large-distortion impact simulations. The fragility estimate for risk assessments uses strain energy concepts, a ductile-rupture damage criterion and tri-linear stress-strain curves adjusted for material cold-forming triaxiality and weldment toughness scatter. FEA examples are provided for waste package impacts on ground support structures.
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Beceiro, Alvaro R., Elena Vico i Emilio G. Neri. "The Radioactive Waste Management Programme in Spain". W ASME 2003 9th International Conference on Radioactive Waste Management and Environmental Remediation. ASMEDC, 2003. http://dx.doi.org/10.1115/icem2003-4898.
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The paper will start with an overview of the organisational and financing framework established in Spain for the safe and efficient management of radioactive waste and decommissioning of nuclear installations. Since its creation by Royal Decree in 1984, ENRESA, the Spanish Radioactive Waste Management Agency, is in charge of both activities. ENRESA is a state owned company whose shareholders are CIEMAT (Centro de Investigaciones Energe´ticas, Medioambientales y Tecnolo´gicas) and the State Industrial Holding (SEPI), both governmental institutions. In Spain the Directorate General for Energy Policy and Mines of the Ministry of Economy (MINECO) plays the leading role in controlling nuclear activities, since it is the body responsible for awarding licenses and permits for installations and activities within the framework of the existing nuclear legislation. The Nuclear Safety Council (CSN) was set up in 1980 as the only competent body in matters of nuclear safety and radiological protection, and is generally responsible for the regulation and supervision of nuclear installations. Any license granted by MINECO is subjected to the mandatory and binding report of the CSN. The paper will review the steps undertaken for solving the national problems associated with the management of radioactive waste and decommissioning of nuclear installations, including uranium mining and milling facilities, and will address the lessons learnt from the activities developed by ENRESA and the future goals to be met. Regarding the L/ILW (Low and Intermediate Level Radioactive Waste) programme, the main milestones of El Cabril L/ILW disposal facility will be described highlighting the most relevant events as well as the foreseen activities, most of them focus on optimizing the capacity of the already operating installation. The elaboration and signature of a Protocol, at the end of 1999, for collaboration on the radiological Surveillance of Metallic Materials in order to detect the possible presence of radioactive materials is worth to be mentioned because of the involvement and agreement of several public and private organisations as well as the administration. Concerning the SF and HLW (Spent Nuclear Fuel and High Level radioactive Waste) programme, the solutions adopted in order to solve the insufficient capacity of the storage pools at NPPs will be described as well as the evolution of the final disposal programme since its beginning and the foreseen goals to be achieved before the year 2010. The last activities will deal with the experience gained during the decommissioning of Vandello´s I NPP and the future decommissioning projects. The decision taken in 2002 by the Spanish authorities to close down Jose´ Cabrera NPP in April 2006, before its 40 years lifetime, has had an impact on ENRESA’s activities.
Raporty organizacyjne na temat "Metallic nuclear waste"
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McDeavitt, S. M., D. P. Abraham, J. Y. Park i D. D. Jr Keiser. Stainless steel-zirconium alloy waste forms for metallic fission products and actinides during treatment of spent nuclear fuel. Office of Scientific and Technical Information (OSTI), lipiec 1996. http://dx.doi.org/10.2172/270551.
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