Thematische Bibliographien / Nuclear decommissioning and dismantling

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Auswahl der wissenschaftlichen Literatur zum Thema „Nuclear decommissioning and dismantling“

Autor: Grafiati

Veröffentlicht am 25. Mai 2024

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Zeitschriftenartikel zum Thema "Nuclear decommissioning and dismantling"

Kim, Hyeon-Ki, Sang-Hwa Shin, Chang-Sig Kong und Chang-Lak Kim. „Evaluation of Worker Radiation Exposure during the Kori Unit 1 Steam Generator Dismantling Process“. Science and Technology of Nuclear Installations 2024 (30.01.2024): 1–7. http://dx.doi.org/10.1155/2024/4230293.

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Kori Unit 1 was permanently shut down on June 18, 2017. Since then, Korea is actively preparing for the decommissioning of the nuclear power plant. Because decommissioning work is performed in a radioactive environment, worker radiation exposure is a significant consideration. In this study, worker radiation exposure is evaluated during the steam generator, one of the heavy components of nuclear power plant, dismantling process. A radiation evaluation for the dismantling process is performed using the code RESRAD-BUILD. A steam generator dismantling scenario and optimal cutting method are designed to evaluate worker radiation exposure, considering pipe dimensions, cutting tool speed, and experience in steam generator replacement. The evaluation results are derived for each work type and year. As a result of the evaluation, worker radiation exposure is 7.5 man-mSv at the year of planned decommissioning.

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Dragusin, Mitica, Octavian Pavelescu und Ioan Iorga. „Good practices in decommissioning planning and pre-decommissioning activities for the Magurele VVR-S nuclear research reactor“. Nuclear Technology and Radiation Protection 26, Nr. 1 (2011): 84–91. http://dx.doi.org/10.2298/ntrp1101084d.

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The VVR-S Nuclear Research Reactor at the ?Horia Hulubei? National Institute of Physics and Nuclear Engineering in Magurele, Bucharest, will be decommissioned applying the immediate dismantling strategy. The implementation of the decommissioning project started in 2010 and is planned for completion within 11 years. Good practices in decommissioning planning, organization, funding, and logistics are described in this paper.

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Babilas, Egidijus, Eugenijus Ušpuras, Sigitas Rimkevičius, Gintautas Dundulis und Mindaugas Vaišnoras. „Safety Assessment of Low-Contaminated Equipment Dismantling at Nuclear Power Plants“. Science and Technology of Nuclear Installations 2015 (2015): 1–11. http://dx.doi.org/10.1155/2015/650810.

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The decommissioning of nuclear facilities requires adequate planning and demonstration that dismantling and decontamination activities can be conducted safely. Existing safety standards require that an appropriate safety assessment be performed to support the decommissioning plan for each facility (International Atomic Energy Agency, 2006). This paper presents safety assessment approach used in Lithuania during the development of the first dismantling and decontamination project for Ignalina NPP. The paper will mainly focus on the identification and assessment of the hazards raised due to dismantling and decontamination activities at Ignalina Nuclear Power Plant and on the assessment of the nonradiological and radiological consequences of the indicated most dangerous initiating event. The drop of heavy item was indicated as one of most dangerous initiating events for the discussed Ignalina Nuclear Power Plant dismantling and decontamination project. For the analysis of the nonradiological impact the finite element model for the load drop force calculation was developed. The radiological impact was evaluated in those accident cases which would lead to the worst radiological consequences. The assessments results show that structural integrity of the building and supporting columns of building structures will be maintained and radiological consequences are lower than the annual regulatory operator dose limit.

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Ilyasov, Damir Fatovich, Artem Yurievich Ivanov, Nikita Petrovich Agafonov, Anastasiya Andreevna Mikhailenko, Ilya Dmitrievich Ovchinnikov und Polina Olegovna Stepanyan. „Software Development for the Nuclear and Radiation Hazardous Objects Elimination Projects Cost Estimating Using Digital Modeling“. Теоретическая и прикладная экономика, Nr. 4 (April 2022): 67–79. http://dx.doi.org/10.25136/2409-8647.2022.4.38996.

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This article discusses the problems of estimating the projects cost for decommissioning nuclear and radiation hazardous facilities based on BIM data. The software developed at Nuclear Safety Institute of the Russian Academy of Sciences for planning and analyzing decommissioning facilities at the pre-project stage processes is described. In particular, the software main functions are demonstrated: evaluation of the dismantling and decontamination works cost, forecasting the waste generated volume, technological processes planning and the safe waste management cost evaluation, results analysis taking into account the uncertainty of the initial data and sensitivity analysis. The scientific novelty consists in the development by a team of authors of software for financial and economic planning of decommissioning of nuclear and radiation hazardous facilities on the basis of digital information 3D models of objects being created. The need for such development is conditioned by the requirements for systematization and analysis of data on nuclear waste at the preparatory stage for the selection of effective technologies for dismantling and decontamination works and management of radioactive waste, as well as to improve the efficiency of individual decommissioning projects and the Federal Target Program "Providing Nuclear and Radiation Safety for 2016-2020 and for the period up to year 2035" generally.

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Lobach, Yu M., S. Yu Lobach und V. M. Shevel. „Preliminary safety analysis at the decommissioning of the WWR-M research reactor“. Nuclear Physics and Atomic Energy 23, Nr. 2 (25.06.2022): 107–15. http://dx.doi.org/10.15407/jnpae2022.02.107.

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Following the demands established by the current Ukrainian legislation, the Decommissioning Concept for the WWR-M research reactor was recently approved. The Concept envisages a strategy of immediate dismantling; it identifies and justifies the main technical and organizational measures for the preparation and implementation of decommissioning, the sequence of planned works and activities, as well as the necessary conditions and infrastructure. Decommissioning requires proper planning and demonstration that all planned dismantling works will be carried out safely. Presented safety assessment is a mandatory component of the Concept and the most important element of the overarching technological scheme. The purpose of the safety analysis is to provide input for detailed planning on how to ensure safety during decommissioning. Based on the results of the safety analysis, the measures to ensure radiation protection are defined while justifying their necessity and sufficiency.

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Lobach, Yu M., S. Yu Lobach, E. D. Luferenko und V. M. Shevel. „Assessment of the dose load during the dismantling of the WWR-M reactor“. Nuclear Physics and Atomic Energy 23, Nr. 4 (25.12.2022): 234–44. http://dx.doi.org/10.15407/jnpae2022.04.234.

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The WWR-M is a light-water-cooled and moderated heterogeneous research reactor with a thermal output of 10 MW. The final decommissioning planning is in progress now. The general decommissioning strategy consists of the dismantling and separate removal of the bulky elements as a whole (in one piece) without preliminary segmentation. The dismantling of the primary and secondary cooling loops is considered as one of the key tasks; a separate dismantling design has been developed. The baseline principles for the technical solution and safety are presented in the given paper. Results of the dose assessment showed that the work can be performed at a collective dose of less than 20 man-mSv.

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Mednikov, I. V., V. V. Vasilyev, A. S. Busygin und 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, Nr. 1 (31.03.2020): 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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Engovatov, Igor A., und Rinat Kh Adiyatullin. „Providing rationale for the possibility of decommissioning Bilibino nuclear cogeneration plant based on the onsite disposal option“. Nuclear Energy and Technology 6, Nr. 3 (06.11.2020): 195–201. http://dx.doi.org/10.3897/nucet.6.58969.

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The problem of the NPP decommissioning after the end of the specified or extended life has reached the practical solution stage for countries possessing a nuclear power industry. The major decommissioning options, both in Russia and abroad, include immediate dismantling and deferred dismantling. At the same time, there are NPP units for which, for a number of reasons, none of the two options are acceptable in terms of ensuring the safety of the personnel, the public and the environment. Disposal, the third and a more rare option, shall be used for decommissioning in this case. The purpose of the work is to provide rationale for the possibility of decommissioning Bilibino Nuclear Cogeneration Plant based on the Onsite Disposal option by covering the main building with an inert material with the formation of a mound. The option has been selected considering the results of an integrated analysis taking into account the geographical, operational, radiological, and socioeconomic factors, as well as based on a limited experience of decommissioning commercial uranium-graphite reactors both within and outside Russia. In accordance with Russian law, the decommissioning stage will start after spent nuclear fuel is withdrawn from the unit and removed. Emphasis is placed on the proposed option preparation and implementation issues. Dates and sequences for the performance of operations to dismantle the components and civil works of buildings and structures, as well as the onsite protective mound formation structure and composition are discussed. The geometrical dimensions, as well as the quantities and types of the mound-forming materials have been estimated. The key mound-forming materials will be fragments of the components, the biological shielding, and the civil works, as well as local materials.

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Delgado, Jessica C., Felix Pino, Erica Fanchini, Alessandro Iovane, Daniela Fabris und Sandra Moretto. „Neutron-gamma survey system for decommissioning and dismantling activities“. EPJ Web of Conferences 288 (2023): 07010. http://dx.doi.org/10.1051/epjconf/202328807010.

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The nuclear plant decommissioning and dismantling (D&D) operations will amount to €200 billion in costs over decades around the world, with three-fourths coming from Europe. Decommissioning includes activities such as planning, physical and radiological characterization, facility and site decontamination, dismantling, and materials management. This work is focused on the development of a compact, light and low-power consumption neutron-gamma survey system which could be easily mounted on an remotely operated vehicle. It is made up of a 4”x4”x2” NaIL (NaI:Tl + 1% 6Li [95% enriched]) neutron/gamma scintillation detector coupled to a SiPM array. Digital pulse processing techniques were implemented to acquire and process the signals, by means of a CAEN DT5780 unit. A comprehensive characterization of this system, based on experiments and Monte Carlo simulations, is reported. The system can be used as a secondary inspection tool, useful for identifying radioactive and special nuclear materials in hotspots.

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Craig, David, Lorna Fecitt, Yuri Gorlinsky, Neil Harman, Roger Jackson, Vyacheslav Kolyadin, Yuri Lobach und Vitaly Pavlenko. „Technical features of the MR reactor decommissioning“. Nuclear Technology and Radiation Protection 23, Nr. 2 (2008): 79–85. http://dx.doi.org/10.2298/ntrp0802079c.

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This paper presents a preliminary technical design for the dismantling of the MR reactor. The goal of the design is the removal of reactor components allowing the re-use of the building for a different nuclear related purpose. The sequence of segmentation procedures is established. Considerations on the size reduction and tooling are presented.

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Dissertationen zum Thema "Nuclear decommissioning and dismantling"

Achigar, Sophie. „Vitrification de déchets nucléaires de démantèlement riches en Mo, P et Zr. Etude structurale et microstructurale de leur incorporation dans un verre aluminoborosilicaté“. Electronic Thesis or Diss., Université Paris sciences et lettres, 2020. http://www.theses.fr/2020UPSLC019.

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Ce travail de thèse s’inscrit dans le projet DEM’N’MELT dont le but est de vitrifier des déchets de moyenne ou haute activité issus du démantèlement d’installations nucléaires. Les compositions de déchet considérées dans ce manuscrit, riches en P2O5, MoO3 et ZrO2 et dont l’activité résulte essentiellement du 137Cs, sont basées sur celles des déchets générés par le démantèlement de l’usine UP1 de Marcoule. Leur principale caractéristique est leur variabilité de composition. L’objectif est d’étudier l’incorporation de ces déchets dans un verre aluminoborosilicaté riche en alcalins à 1100 °C.Le premier axe d’étude consiste à se placer dans un système proche du système industriel (11 oxydes). Il a mis en évidence que MoO3 et P2O5 sont les deux principaux constituants du déchet conduisant à des séparations de phases et/ou des cristallisations. Celles-ci peuvent, dans le cas des phases molybdates, contenir du Cs. Aux teneurs envisagées, ZrO2 s’incorpore quant à lui dans la matrice sans générer d’hétérogénéités.Le deuxième axe se concentre sur l’étude structurale et microstructurale des mécanismes d’incorporation de P2O5 et MoO3 dans un système simplifié (6-7 oxydes). Ces éléments sont tout d’abord considérés seuls puis incorporés conjointement. Il apparaît que P et Mo s’insèrent majoritairement sous forme d’entités isolées (PO43- et MoO42-) du réseau vitreux et que leur incorporation conjointe augmente la tendance à la cristallisation du système
This work belongs to the DEM’N’MELT project, which is dedicated to the vitrification of intermediate or high level radioactive wastes coming from the dismantling of nuclear facilities. The waste compositions of this study, rich in P2O5, MoO3 et ZrO2 which activity is mainly due to 137Cs are close to the ones of the shutdown UP1 facility (Marcoule). Their main feature is the variability of their composition. This work objective is to study the incorporation of these wastes in an aluminoborosilicate glass rich in alkali oxides at 1100 °C.The first part of the study will be dedicated to a system close to the industrial one (11 oxides). It highlights that MoO3 and P2O5 are the main waste constituents responsible for phase separation or crystallization. Moreover, molybdate crystalline phases can contain Cs. ZrO2 is incorporated in the glassy matrix without leading to heterogeneities.Then, a simplified system (6-7 oxides) is studied along with the structural and microstructural incorporation mecanisms of P2O5 and MoO3. These oxides are first considered alone and then added simultaneously. This second study highlights that P et Mo mainly lead to the formation of entities isolated from the glassy network and that their simultaneous addition increases the crystallization tendency

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Kim, Dae Ji. „Tritium speciation in nuclear decommissioning materials“. Thesis, University of Southampton, 2009. https://eprints.soton.ac.uk/72145/.

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Tritium is a by-product of civil nuclear reactors, military nuclear applications, fusion programmes and radiopharmaceutical production. It commonly occurs, though not exclusively, as tritiated water (HTO) or organically-bound tritium (OBT) in the environment but may exist as other forms in nuclear-related construction and fabrication materials. During the lifetime of nuclear sites (especially those involving heavy water) tritium becomes variably incorporated into the fabric of the buildings. When nuclear decommissioning works and environmental assessments are undertaken it is necessary to accurately evaluate tritium activities in a wide range of materials prior to any waste sentencing. Of the various materials comprising UK radioactive wastes, concrete and metal account for approximately 20% of the total weight of low level waste (LLW) and 12% and 35% of the total weight of intermediate level waste (ILW). Proper sampling and storage of samples are significant factors in achieving accurate tritium activities. The degree of loss of 3H and cross-contamination can be significantly reduced by storing samples in an air/water tight container in a freezer (-18°C). The potential for tritium contamination is dependent on the 3H form. Most 3H loss originates from tritiated water which is easily exchanged with atmospheric hydrogen in the form of water vapour at room temperature. However, the loss of more strongly bound 3H, produced in-situ in materials by neutron activation, is not significant even at room temperature. Such tritium is tightly retained in materials and does not readily exchange with water or diffuse. In nuclear reactor environments tritium may be produced via several neutron-induced reactions, 2H(n,g)3H, 6Li(n,a)3H, 10B(n,2a)3H and ternary fission (fission yield <0.01%). It may also exist as tritiated water (HTO) that is able to migrate readily and can adsorb onto various construction materials such as structural concrete. In such locations it exists as a weakly-bound form that can be lost at ambient temperatures. Bioshield concretes present a special case and systematic analysis of a sequence of sub-samples taken from a bioshield core (from UKAEA Winfrith) has identified a strongly-bound form of 3H in addition to the weakly bound form. The strongly bound 3H in concrete is held more strongly in mineral lattices and requires a temperature of >850°C to achieve quantitative recovery. This more strongly retained tritium originates from neutron capture of trace lithium (6Li and potentially 10B) distributed throughout minerals in the concrete. The highest proportion of strongly bound 3H was observed in the core sections closest to the core. Weakly bound tritium is associated with water loss from hydrated mineral components. Tritium is retained in metals by absorption by free water, hydrated surface oxidation layer, H ingress into bulk metal and also as lattice-bound tritium produced via in-situ neutron activation. Away from the possible influence of neutrons, the main 3H contamination to metals arises from absorption and diffusion via atmospheric exposure to the HTO. Here contamination is mainly confined to the metal surface layer. The tritium penetration rate into metal surfaces is controlled by the metal type and its surface condition. Where metals are exposed to a significant neutron flux and contain 6Li, 7Li and 10B then in situ 3H production will occur which may propagate beyond the surface layer. In such cases tritium may exist in two forms namely a weakly bound HTO form and a non-HTO strongly bound form. The HTO form is readily lost at moderate temperatures (~120°C) whereas the non-HTO requires up to 850°C for complete extraction.

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McBryde, Daniel John. „Ice pigging in the nuclear decommissioning industry“. Thesis, University of Bristol, 2015. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.702749.

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Ice pigging is a novel technology using thick ice slurry (a two-phase mixture of ice crystals and freezing point depressant solution) to clean the internal surfaces of pipes or ducts; this mixture displays semi-solid characteristics. When pumped through a pipe, the slurry adopts plug flow, forming an 'ice pig'; slip occurs at the interface with the pipe walls generating high shear stresses; thus able to mobilise and remove sediment residing at the pipe wall. Ice pigs are able to navigate demanding topologies such as vertical falls, diameter changes, orifice plates, heat exchangers, and intrusive instrumentation; they provide a method of removing fouling without the need for dis-assembly, reducing valuable down-time, labour intensive pipe work dismantling, and subsequent manual cleaning. Many decades of nuclear activity here in the UK have produced unique and difficult challenges that require solving at Sellafield, the UK's nuclear waste reprocessing site. The drive to produce plutonium for atomic weapons during the 1950's, with very little foresight towards how the wastes and facilities would be dealt with, has brought about significant challenges. As these facilities are nearing the end of their design lives, the time has come to assess methods of treating these wastes and decommissioning the facilities in a safe, controlled, and cost-effective manner. Ice pigging is one of many technologies being assessed for such a task; this thesis details specific areas of application where experimental work has been conducted. Experimental work conducted in this thesis has: developed a method of characterising the ice pig's sediment removal performance compared to simple water flushing, assessed the ice pig's ability to remove representative sediments, assessed the ice pig's suitability for removing sediment from heat exchangers to restore thermal performance, and analysed the rate of percolation of the driving fluid through the ice pig body, such that the suitability of the ice pig for separating fluids can be established.

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Fourtakas, Georgios. „Modelling multi-phase flows in nuclear decommissioning using SPH“. Thesis, University of Manchester, 2014. https://www.research.manchester.ac.uk/portal/en/theses/modelling-multiphase-flows-in-nuclear-decommissioning-using-sph(f5ed0b5b-ea62-431a-bb6e-a18635d396bc).html.

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This thesis presents a two-phase liquid-solid numerical model using Smoothed Particle Hydrodynamics (SPH). The scheme is developed for multi-phase flows in industrial tanks containing sediment used in the nuclear industry for decommissioning. These two-phase liquid-sediments flows feature a changing interfacial profile, large deformations and fragmentation of the interface with internal jets generating resuspension of the solid phase. SPH is a meshless Lagrangian discretization scheme whose major advantage is the absence of a mesh making the method ideal for interfacial and highly non-linear flows with fragmentation and resuspension. Emphasis has been given to the yield profile and rheological characteristics of the sediment solid phase using a yielding, shear and suspension layer which is needed to predict accurately the erosion phenomena. The numerical SPH scheme is based on the explicit treatment of both phases using Newtonian and non-Newtonian Bingham-type constitutive models. This is supplemented by a yield criterion to predict the onset of yielding of the sediment surface and a suspension model at low volumetric concentrations of sediment solid. The multi-phase model has been compared with experimental and 2-D reference numerical models for scour following a dry-bed dam break yielding satisfactory results and improvements over well-known SPH multi-phase models. A 3-D case using more than 4 million particles, that is to the author’s best knowledge one of the largest liquid-sediment SPH simulations, is presented for the first time. The numerical model is accelerated with the use of Graphic Processing Units (GPUs), with massively parallel capabilities. With the adoption of a multi-phase model the computational requirements increase due to extra arithmetic operations required to resolve both phases and the additional memory requirements for storing a second phase in the device memory. The open source weakly compressible SPH solver DualSPHysics was chosen as the platform for both CPU and GPU implementations. The implementation and optimisation of the multi-phase GPU code achieved a speed up of over 50 compared to a single thread serial code. Prior to this thesis, large resolution liquid-solid simulations were prohibitive and 3-D simulations with millions of particles were unfeasible unless variable particle resolution was employed. Finally, the thesis addresses the challenging problem of enforcing wall boundary conditions in SPH with a novel extension of an existing Modified Virtual Boundary Particle (MVBP) technique. In contrast to the MVBP method, the extended MVBP (eMVBP) boundary condition guarantees that arbitrarily complex domains can be readily discretized ensuring approximate zeroth and first order consistency for all particles whose smoothing kernel support overlaps the boundary. The 2-D eMVBP method has also been extended to 3-D using boundary surfaces discretized into sets of triangular planes to represent the solid wall. Boundary particles are then obtained by translating a full uniform stencil according to the fluid particle position and applying an efficient ray casting algorithm to select particles inside the fluid domain. No special treatment for corners and low computational cost make the method ideal for GPU parallelization. The models are validated for a number of 2-D and 3-D cases, where significantly improved behaviour is obtained in comparison with the conventional boundary techniques. Finally the capability of the numerical scheme to simulate a dam break simulation is also shown in 2-D and 3-D.

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Fort, Emily Minatra. „A historical site assessment of the Georgia Tech Research Reactor“. Thesis, Georgia Institute of Technology, 1999. http://hdl.handle.net/1853/17257.

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Grabrovaz, Meaghan. „An investigation into the forecasting of skills in nuclear decommissioning“. Thesis, University of Central Lancashire, 2017. http://clok.uclan.ac.uk/23759/.

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This study explores the nature of skills forecasting in nuclear decommissioning and that which makes skills forecasting information useful. The study adopts a pragmatic approach using an interpretative, qualitative case study research design and draws on aspects of a critical realist approach to uncover, deconstruct and challenge some ‘norms’ in skills forecasting. The study makes an original contribution to knowledge through the identification of nineteen factors that influence skills forecasting in the nuclear industry. It also generates a baseline of knowledge on the theory and practice of skills forecasting and management through a review of the literature on skills, forecasting, skills forecasting and workforce planning and relevant aspects of public sector management and HRM. The study documents and compares current skills forecasting practice amongst UK site licensed companies and selected supply chain companies. Such research has not previously been conducted in the nuclear decommissioning industry. This answers research questions about why, and how, different groups in the sector perform skills forecasting and how variations in approaches affect the information produced. It also answers research questions about who uses skills forecasting information, and how. Together with a review of current problems with skills information, this contributes to an understanding of what makes skills information useful. The research evidences that while the industry has some common features with other High Reliability Organisations, there are unique dimensions which make this research significant. Some ‘norms’ operating in skills forecasting were challenged including how it is being used, eg as an agent for change by some groups, and assumptions about the potential availability of skills from the supply chain. The literature review was used to construct a practical-ideal type, an approach derived from classical pragmatism offering a version of a nearly ideal process, on the understanding that this is socially constructed and subject to continual change. Existing practice is evaluated against this practical-ideal type in a unique application of this methodology in the nuclear decommissioning context.

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Nancekievill, Matthew. „The radiation tolerance and development of robotic platforms for nuclear decommissioning“. Thesis, University of Manchester, 2018. https://www.research.manchester.ac.uk/portal/en/theses/the-radiation-tolerance-and-development-of-robotic-platforms-for-nuclear-decommissioning(75451a19-57c6-4809-92dd-9b683db9b10f).html.

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There is an increasing desire to deploy low-cost robotic systems in nuclear decommissioning environments. These environments include long-standing nuclear fuel storage ponds such as those at the Sellafield site in Cumbria, UK as well as areas affected by expulsion of radioactive material from sites such as the Fukushima accident in Japan 2011. An area of concern for the successful deployment of robotic platforms in a radioactive field is their radiation tolerance. It is necessary to understand how the low-cost components used within robotic platforms react to radiation exposure in a nuclear decommissioning environment. This thesis discusses the radiation tolerance of multiple commercial-off-the-shelf (COTS) components that are commonly used within a robotic platform up to an expected yearly total dose of 5 kGy(Si). It was found that COTS voltage regulators are susceptible to gamma exposure, however, development of a discrete voltage regulator showed an increased tolerance to radiation under certain load and temperature conditions. Inertial measurement units were also investigated and found to be susceptible to a total ionising dose.

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Dallimore, Matthew. „Gamma ray imaging in industrial and medical applications“. Thesis, University of Southampton, 2001. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.246854.

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Snell, Benjamin Aaron. „Dismantling Russia's Northern Fleet Nuclear Submarines environmental and proliferation risks /“. Thesis, Monterey, Calif. : Springfield, Va. : Naval Postgraduate School ; Available from National Technical Information Service, 2000. http://handle.dtic.mil/100.2/ADA378654.

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Thesis (M.A. in National Security Affairs) Naval Postgraduate School, June 2000.
Thesis advisor(s): Yost, David S.; Minott, Rodney K. "June 2000." Includes bibliographical references. Also available online.

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LING, XIANBING. „BAYESIAN ANALYSIS FOR THE SITE-SPECIFIC DOSE MODELING IN NUCLEAR POWER PLANT DECOMMISSIONING“. NCSU, 2001. http://www.lib.ncsu.edu/theses/available/etd-20010130-141644.

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Decommissioning is the process of closing down a facility. In nuclear power plant decommissioning, it must be determined that that any remaining radioactivity at a decommissioned site will not pose unacceptable risk to any member of the public after the release of the site. This is demonstrated by the use of predictive computer models for dose assessment. The objective of this thesis is to demonstrate the methodologies of site-specific dose assessment with the use of Bayesian analysis for nuclear power plant decommissioning. An actual decommissioning plant site is used as a test case for the analyses. A residential farmer scenario was used in the analysis with the two of the most common computer codes for dose assessment, i.e., DandD and RESRAD. By identifying key radionuclides and parameters of importance in dose assessment for the site conceptual model, available data on these parameters was identified (as prior information) from the existing default input data from the computer codes or the national database. The site-specific data were developed using the results of field investigations at the site, historical records at the site, regional database, and the relevant information from the literature. This new data were compared to the prior information with respect to their impacts onboth deterministic and probabilistic dose assessment. Then, the two sets of information were combined by using the method of conjugate-pair for Bayesian updating. Value of information (VOI) analysis was also performed based on the results of dose assessment for different radionuclides and parameters. The results of VOI analysis indicated that the value of site-specific information was very low regarding the decision on site release. This observation was held for both of the computer codes used. Although the value of new information was very low with regards to the decisions on site release, it was also found that the use of site-specific information is very important for the reduction of the predicted dose. This would be particularly true with the DandD code.

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Bücher zum Thema "Nuclear decommissioning and dismantling"

Agency, International Atomic Energy, Hrsg. Dismantling of contaminated stacks at nuclear facilities. Vienna: International Atomic Energy Agency, 2005.

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Sarkisov, Ashot A., und Alain Tournyol Clos, Hrsg. Analysis of Risks Associated with Nuclear Submarine Decommissioning, Dismantling and Disposal. Dordrecht: Springer Netherlands, 1999. http://dx.doi.org/10.1007/978-94-011-4595-4.

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A, Sarkisov A., Tournyol Du Clos Alain und NATO Advanced Research Workshop on Analysis of Risks Associated with Nuclear Submarine Decommissioning, Dismantling, and Disposal (1997 : Moscow, Russia), Hrsg. Analysis of risks associated with nuclear submarine decommissioning, dismantling, and disposal. Dordrecht: Kluwer, 1999.

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Great Britain. Health and Safety Commission. Environmental impact assessment: Dismantling and decommissioning nuclear reactors and power stations. [U.K.]: Health and Safety Executive, 1998.

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Sarkisov, Ashot A. Analysis of Risks Associated with Nuclear Submarine Decommissioning, Dismantling and Disposal. Dordrecht: Springer Netherlands, 1999.

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NATO Advanced Research Workshop on Analysis of Risks Associated with Nuclear Submarine Decommissioning, Dismantling and Disposal (1997 Moscow, Russia, Federation). Analysis of risks associated with nuclear submarine decommissioning, dismantling and disposal: [proceedings of the NATO Research Advanced Workshop on Analysis of Risks Associated with Nuclear Submarine Decommissioning, Dismantling and Disposal, Moscow, Russia, November 24-26, 1997]. Dordrecht: Kluwer Academic, 1999.

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Laraia, Michele. Nuclear Decommissioning. Cham: Springer International Publishing, 2018. http://dx.doi.org/10.1007/978-3-319-75916-6.

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Taking forward decommissioning: The Nuclear Decommissioning Authority. London: Stationery Office, 2008.

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K, Pflugrad, und Commission of the European Communities. Directorate-General for Science, Research, and Development., Hrsg. Decommissioning of nuclear installations. London: Elsevier Applied Science, 1990.

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Decommissioning: Nuclear power's missing link. Washington, D.C: Worldwatch Institute, 1986.

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Buchteile zum Thema "Nuclear decommissioning and dismantling"

Díaz-Díaz, J. L. „Decommissioning of Nuclear Installations and Dismantling Techniques“. In The Environmental Challenges of Nuclear Disarmament, 125–30. Dordrecht: Springer Netherlands, 2000. http://dx.doi.org/10.1007/978-94-011-4104-8_14.

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Jackson, Philip K. „Risks Encountered in Nuclear Decommissioning and Their Applicability to Nuclear Submarine Decommissioning“. In Analysis of Risks Associated with Nuclear Submarine Decommissioning, Dismantling and Disposal, 155–64. Dordrecht: Springer Netherlands, 1999. http://dx.doi.org/10.1007/978-94-011-4595-4_20.

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Nikishin, G. D., V. P. Sharikov, O. G. Sokolov und V. V. Yurchenko. „Ensuring of Radiation Safety when Decommissioning, Dismantling and Recycling Nuclear Submarines“. In Nuclear Submarine Decommissioning and Related Problems, 273–76. Dordrecht: Springer Netherlands, 1996. http://dx.doi.org/10.1007/978-94-009-1758-3_34.

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Testov, S. G., Ye V. Tikhomirov und V. M. Zakharov. „Top-Priority Issues of Radiation Safety in Decommissioning, Dismantling, and Recycling Nuclear Submarines“. In Nuclear Submarine Decommissioning and Related Problems, 269–71. Dordrecht: Springer Netherlands, 1996. http://dx.doi.org/10.1007/978-94-009-1758-3_33.

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Robin, Bernard, und Amaury Buzonniere. „Decommissioning and Dismantling of French Nuclear Submarines: Industral Issues“. In Remaining Issues in the Decommissioning of Nuclear Powered Vessels, 63–64. Dordrecht: Springer Netherlands, 2003. http://dx.doi.org/10.1007/978-94-010-0209-7_8.

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Ølgaard, Povl L. „Nuclear Risks of Decommissioned Nuclear Submarines with Non-Defueled Reactors“. In Analysis of Risks Associated with Nuclear Submarine Decommissioning, Dismantling and Disposal, 87–94. Dordrecht: Springer Netherlands, 1999. http://dx.doi.org/10.1007/978-94-011-4595-4_13.

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Barskov, M. K., S. D. Gavrilov, P. L. Smirnov, V. P. Shcherbak und N. N. Yurasov. „Nuclear Fuel Cycle for Russian Ship Spent Nuclear Fuel: Reprocessing or Direct Disposal?“ In Analysis of Risks Associated with Nuclear Submarine Decommissioning, Dismantling and Disposal, 303–11. Dordrecht: Springer Netherlands, 1999. http://dx.doi.org/10.1007/978-94-011-4595-4_35.

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Vysotsky, V. L., und V. A. Danilyan. „Radioecological Hazards from Decommissioning and Recycling of Nuclear Propelled Ships“. In Analysis of Risks Associated with Nuclear Submarine Decommissioning, Dismantling and Disposal, 331–37. Dordrecht: Springer Netherlands, 1999. http://dx.doi.org/10.1007/978-94-011-4595-4_39.

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Dadoumont, J., V. Massaut und F. Vermeersch. „The BR3 Decommissioning project: A Pilot for Submarine Reactors“. In Analysis of Risks Associated with Nuclear Submarine Decommissioning, Dismantling and Disposal, 103–17. Dordrecht: Springer Netherlands, 1999. http://dx.doi.org/10.1007/978-94-011-4595-4_15.

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Yurasov, N. N., und S. G. Testov. „Normative-Legal Aspects of Nuclear and Radiation Safety Provision during Nuclear Submarine Decommissioning, Storage and Utilization“. In Analysis of Risks Associated with Nuclear Submarine Decommissioning, Dismantling and Disposal, 23–26. Dordrecht: Springer Netherlands, 1999. http://dx.doi.org/10.1007/978-94-011-4595-4_5.

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Konferenzberichte zum Thema "Nuclear decommissioning and dismantling"

Bienia, Harald. „Thermal Cutting Technologies for Decommissioning of Nuclear Facilities“. In ASME 2009 12th International Conference on Environmental Remediation and Radioactive Waste Management. ASMEDC, 2009. http://dx.doi.org/10.1115/icem2009-16297.

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Remote disassembly of radiologically burdened large components is among the most sophisticated and complex activities in the dismantling of nuclear installations. High local dose rates and contamination levels, combined with complicated designs and geometries of the object to be dismantled, plus insufficient accessibility, imply major challenges in the dismantling of nuclear facilities. Usually the shielding effect of water is used during the dismantling period. Other dismantling activities require dry ambiences.

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Iguchi, Yukihiro, Tsuyoshi Tajiri und Shikou Kiyota. „Preparatory Activities of the Fugen Decommissioning“. In 12th International Conference on Nuclear Engineering. ASMEDC, 2004. http://dx.doi.org/10.1115/icone12-49069.

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Advanced Thermal Reactor “Fugen” is a 165MWe, heavy water moderated, light-water cooled, pressure-tube type reactor. In February 1998, the Atomic Energy Commission of Japan decided to end the mission of ATR development and introduced a new policy that development and research of decommissioning of Fugen should be promoted in order to carry out the decommissioning smoothly after the shutdown. The Fugen NPS has shut down permanently in March 2003, and based on the report titled Activities of Fugen NPS after permanent shutdown, “Fugen” has been preparing for the project, including necessary development of technologies. The development of decommissioning for “Fugen” is divided into two areas. One area is developement of unique technology for dismantling special components such as the reactor core and the heavy water system. Another area is improvement and enhancement of existing technologies. Especially the former area requires effort and comprises development of the reactor dismantlement, tritium decontamination of heavy water system and engineering support systems. The concrete activities are as follows: The density and amount of radioactive nuclides in all equipment or concrete including the reactor core need to be evaluated for the decommissioning. To prepare for decommissioning, analysis, measurement and evaluation of the neutron flux density have been executed during reactor operation. Special dismantling process is necessary for the heavy water system and the reactor that are unique to Fugen. Some studies and tests are going on for the safe dismantling based on existing technologies and their combination. It is important to execute the decommissioning economically and rationally. Therefore, systems engineering approach is necessary in order to optimize the work load, exposure dose, waste mass and cost by selecting appropriate dismantling process at the planning stage of the decommissioning. For this reason, in order to make a decommissioning plan efficiently, we have been developing an Engineering Support System for decommissioning by adopting new information technologies such as three-dimensional- (3D-) CAD system and virtual reality (VR) system. Moreover, the technical results of technology development and decommissioning activity should be organized and opened to the public so that they can contribute to other decommissioning projects. For this purpose, information exchange and co-operation with domestic and international organizations are underway.

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Vidaechea, Sergio, und Manuel Ondaro. „The Decommissioning of the CIEMAT Nuclear Research Center“. In ASME 2011 14th International Conference on Environmental Remediation and Radioactive Waste Management. ASMEDC, 2011. http://dx.doi.org/10.1115/icem2011-59321.

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CIEMAT, formerly the Nuclear Energy Board (JEN), is the Spanish Centre for Energy-Related, Environmental and Technological Research. Located in Madrid, it used to have more than 60 facilities in operation that allowed a wide range of activities in the nuclear field and in the application of ionising radiations. Particularly significant among these facilities were the research reactors, particle accelerators, hot cells and nuclear fuel manufacturing and processing plants. At present the centre, which is authorised as a single nuclear facility, includes various installations, some of them are now obsolete, shut down and in dismantling phases. In 2000 CIEMAT started the “Integrated plan for the improvement of CIEMAT installations (PIMIC)”, which includes activities for the decontamination, dismantling and rehabilitation of obsolete facilities. This paper will describe the decommissioning process carried out at CIEMAT Research Centre, and will review the key aspects of the project, including site remediation and release.

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Shimada, Taro, Soichiro Ohshima und Takenori Sukegawa. „Development of Safety Assessment Code for Decommissioning of Nuclear Facilities (DecDose)“. In 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75123.

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A safety assessment code, DecDose, for decommissioning of nuclear facilities has been developed, based on the experiences of the decommissioning project of Japan Power Demonstration Reactor (JPDR) at Japan Atomic Energy Research Institute (now Japan Atomic Energy Agency). DecDose evaluates the annual exposure dose of the public and workers according to the progress of decommissioning of the plant, and also evaluates the public dose at accidental situations including fire and explosion. The public dose at normal situations during decommissioning is evaluated from the amount of radionuclides discharged from the plant to the atmosphere and the ocean. The amounts of radionuclides discharged depend on which and how activated and/or contaminated components and structures are dismantled. The amount is predicted by using the radioactive inventory given by the plant. The filtration efficiency of the ventilation system and decontamination factors of the liquid waste treatment system of the plant are also considered. Both of the internal dose caused by inhalation and ingestion of agricultural crops and seafood, and the external dose by radioactive aerosols airborne and radioactive deposition at soil surfaces are calculated for all of possible pathways. Also included in the external dose are direct radiation and skyshine radiation from waste containers which are packed and temporarily stored in the in-site building. For external dose of workers, the radiation dose rate from dismantling contaminated components and structures is calculated using the dose rate library which was previously evaluated by a point kernel shielding code. In this condition, radiation sources are regarded to be consisted of two parts; one is a dismantling object of interest, and the other is the sum of surrounding objects. Difference in job type or position is taken into account; workers for cutting are situated closer to a dismantling component, other workers help them at some distance, and the supervisor watches their activities from away. For worker’s internal dose, the radionuclide concentrations in air for individual radionuclides are calculated from a dismantling condition, e.g. cutting speed, cutting length of the dismantling component and exhaust velocity. A calculation model for working time on dismantling was developed using more segmented WBS (work breakdown structure). DecDose was partially verified by comparison with measured the external dose of workers during JPDR Decommissioning Project. The DecDose is expected to contribute to utilities in formulating rational dismantling plans and to the safety authority in estimating conservativeness in safety assessment of licensing application or risk-based regulatory criteria.

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Fichet, Pascal, Anumaija Leskinen, Sylvie Guegan und Florence Goutelard. „Characterization of Beta Emitters for Decommissioning“. In ASME 2013 15th International Conference on Environmental Remediation and Radioactive Waste Management. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icem2013-96087.

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Radioactive waste management is nowadays, after nearly 50 years of concern, a technical and economical challenge faced by existing nuclear power countries. In decommissioning of nuclear facilities after removal of the nuclear equipment (laboratory materials, glove boxes, etc.), the radioactive inventory of the various building materials is needed to state the working condition for dismantling. Thus, characterization is essential for decommissioning and moreover for radioactive waste classification and management. A radionuclide imaging technique, the Digital Autoradiography (DA), also known as storage phosphor technology, has been studied for decommissioning projects because of its advantages such low price, easy utilization and sensitivity. DA has been proven to be an efficient technic in localization of beta emitter (especially C-14 and H-3) contamination remaining in nuclear facilities under dismantling. Samples have been collected where C-14 or H-3 have been observed by DA and analyzed by the classical technique : pyrolysis followed by liquid scintillation counting. Real applications to classify potential waste coming from a laboratory under dismantling are described in this paper.

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Santiago, Juan L., und Sergio Vidaechea. „Recent Developments in D&D of Nuclear Facilities in Spain“. In ASME 2001 8th International Conference on Radioactive Waste Management and Environmental Remediation. American Society of Mechanical Engineers, 2001. http://dx.doi.org/10.1115/icem2001-1222.

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Abstract Spain occupies a leading position at international level in the field of installation decommissioning. Decommissioning projects have already been performed in relation to uranium mills, the rehabilitation of disused uranium mines is currently in the final phase and the dismantling of the Vandellós-I Nuclear Power Plant is now under way. On the basis of this experience, this paper describes the key issues in decommissioning technology and presents the approaches adopted by ENRESA to tackle the decommissioning strategy in Spain. In particular practical dismantling and decontamination methods are described, and material and radioactive waste management are discussed.

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Vinoche, O., und G. Rodriguez. „Experimental Feedback on Sodium Loop Decommissioning at the CEA“. In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22764.

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The aim of this paper is to present experimental feedback on sodium loop dismantling techniques at the CEA (The French Atomic Energy Commission) and to offer recommendations for the decommissioning of Fast Breeder Reactor secondary sodium loops. This study is divided into several parts which correspond to the different stages of a dismantling system. It is based on acquired CEA decommissioning experience which primarily concerns the following: the decommissioning of RAPSODIE (France’s first Fast Breeder Reactor), the PHENIX reactor secondary loop replacement, the sodium loop decommissioning carried out by the Laboratory of Sodium Technologies and Treatment, and several technical documents. This paper deals with the main results of this survey. First, a comparison of 8 pipe-cutting techniques is made, taking into account speed in cutting, reliability, dissemination, fire risk due to the presence of sodium, cutting depth, and different types of waste (empty pipes, sodium-filled pipes, tanks...). This comparison has led us to recommend the use of an alternative saw or a chain saw rather than the use of the plasma torch or grinder. Different techniques are recommended depending on if they are on-site, initial cuttings or if they are to be carried out in a specially-designed facility referred to hereafter as “the cutting building”. After the cutting stage, the sodium waste must be processed with water to become an ultimate stable waste. Four treatment processes are compared with different standards: speed, cost, low activity adaptability and “large sodium quantity” adaptability. Recommendations are also made for reliable storage, and for the general dismantling system organization. Last, calculations are presented concerning a complete dismantling facility prototype capable of treating large amounts of sodium.

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Ondaro, Manuel. „Jose Cabrera Dismantling and Decommissioning Project“. In ASME 2013 15th International Conference on Environmental Remediation and Radioactive Waste Management. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icem2013-96227.

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The Jose Cabrera Nuclear Power Plant (NPP) was the first commercial power reactor (Westinghouse 1 loop PWR 510 MWth, 160 MWe) commissioned in Spain and provided the base for future development and training. The reactor construction started in 1963 and it was officially on-line by 1969. The NPP operated from 1969 until 2006 when it became the first reactor to be shut down after completing its operational period. The containment is reinforced concrete with a stainless steel head. In 2010 responsibility for D&D was transferred to Enresa to achieve IAEA level 3 (a green field site available for unrestricted re-uses) by 2017. Of the total of more than 104,000 tons of materials that will be generated during dismantling, it is estimated that only ∼ 4,000 tons will be radioactive waste, some of which, 40 t are considered as intermediate level long-lived wastes and the rest (3,960 t) will be categorized as VLLW & ILLW. The Project is divided into five phases: Phase 0 - Removal of fuel and preliminary work. Phase 1 - Preparatory Activities for D&D. complete. Phase 2 - Dismantling of Major Components. Phase 3 - Removal of Auxiliary Installations, Decontamination and Demolition. Phase 4 - Environmental Restoration. Phase 2, is currently ongoing (50% completed). To manage the diverse aspects of decommissioning operations, Enresa uses an internally developed computerized project management tool. The tool, based on knowledge gathered from other Enresa projects, can process operations management, maintenance operations, materials, waste, storage areas, procedures, work permits, operator dose management and records. Enresa considers that communication is important for both internal and external stakeholder relations and can be used to inform, to neutralize negative opinions & attitudes, to remove false expectations and for training. Enresa has created a new multi-purpose area (exhibition/visitor centre) and encourages visits from the public, local schools, local and national politicians and technical groups. Greenfield is the final end state objective. The total cost of this project, including a 20% contingency as estimated in 2003 is 135 M€. This figure does not include the management of the plant spent fuel, which has constituted an independent project that has been completed in 2009 (35 M€). Enresa, with 15 staff on site are managing a team of ∼ 250 workers, 40 of whom belong to the previous operator. The spent fuel is On-Site prior to the final destination in the future Spain Centralized Spent Fuel Installation.

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Koulikov, Konstantin N., Rinat A. Nizamutdinov, Andrey N. Abramov und Anatoly I. Tsubanikov. „Decommissioning and Dismantling Solution Development for Volodarsky Civil Nuclear Fleet Support Ship“. In ASME 2009 12th International Conference on Environmental Remediation and Radioactive Waste Management. ASMEDC, 2009. http://dx.doi.org/10.1115/icem2009-16386.

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Having about 200 tons of solid radioactive waste aboard, the Volodarskiy Floating Technical Base (FTB) is a potential radiation pollution source for the Murmansk region and Kola Bay, as her long-term berthing negatively affects the hull structures. Thereby, Atomflot collaborated with ANO Aspect-Konversia and JSC NIPTB Onega within the frameworks of Federal Special-purpose Program “Assurance of Nuclear and Radiation Safety for 2008 and for the period up to 2015” and developed the Volodarskiy FTB dismantling concept. In 2008 in the course of development of the Volodarskiy FTB dismantling concept the following works were carried out: 1) vessel condition survey, including SRW radiological analysis; 2) feasibility study of the Volodarskiy FTB dismantling alternatives. In this regard the following alternatives were analyzed: – formation of the package assembly in the form of vessel’s undivided hull for durable storage in the Saida long-term storage facility (LTSF); - formation of individual SRW package assemblies for durable storage in the Saida LTSF; - comprehensive recycling of all solid radioactive waste by disposal in protective containers. 3) selection and approval of the dismantling alternative. The alternative of formation of individual SRW package assemblies for durable storage in the Saida LTSF was selected by the Rosatom State Corporation. In this case the works will be performed on a step-by-step basis at the Atomflot enterprise and SRE Nerpa. The conceptual dismantling technology was developed for the selected Volodarskiy FTB dismantling option. The proceedings contain description of options, analysis procedure and proposal for further study of mentioned challenge.

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Kitamura, Akihiro, Takashi Okada, Sinichiro Asazuma, Shinichi Uematsu und Takashi Ishibashi. „Glovebox Dismantling Activities and Decommissioning Plan for Plutonium Fuel Fabricating Facility“. In 12th International Conference on Nuclear Engineering. ASMEDC, 2004. http://dx.doi.org/10.1115/icone12-49511.

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The gloveboxes and process equipment used at plutonium fuel handling facilities have had to be replaced due to deterioration or the need to make changes. So far, their removal and replacement has taken place more than 30 times in Plutonium Fuel Center, Japan Nuclear Cycle Development Institute (JNC). In most recent dismantling activities, we removed four giant gloveboxes (total size, 110 cubic meters) which possessed equipment to recover plutonium from mixed oxide (MOX) fuel scraps. We have implemented a number of procedural improvements in dismantling activities and collected various kinds of data, including type and amount of primary and secondary waste from dismantling, relation between waste volume and work force, etc. Plutonium Fuel Fabricating Facility (PFFF) is one of the three plutonium fuel handling facilities in Plutonium Fuel Center, JNC. Its final mission to produce MOX fuels for the advanced thermal reactor “Fugen” Nuclear Power Station was successfully finished in 2002. Then, we started preparatory activities to draw up a Deactivation & Decommissioning (D&D) plan for this facility and to construct a database with the experimental data of glovebox dismantling activities acquired in the past thirty years. The D&D schedule for this facility can be broken down into three phases. Phase 1 (to 2010): Stabilize all the special nuclear materials in the facility and ship them from the facility. Establish new and effective decontamination and volume reduction technologies in order to improve existing methods. Phase 2 (2010–2015): Apply the abovementioned technologies to some of the glovebox dismantling activities and confirm their adaptability for the project. Draw up a detailed D&D plan which meets to various regulations. Phase 3 (2015–2020): Dismantle all the remaining gloveboxes in the facility and promote research and development of D&D technologies for future projects. Decontaminate inner surfaces of the building in order to reuse the building as a waste storage facility.

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Berichte der Organisationen zum Thema "Nuclear decommissioning and dismantling"

Goodby, James E. Dismantling the Nuclear Weapons Legacy of the Cold War. Fort Belvoir, VA: Defense Technical Information Center, Februar 1995. http://dx.doi.org/10.21236/ada385646.

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Skone, Timothy J. Nuclear Power Plant, Decommissioning. Office of Scientific and Technical Information (OSTI), November 2010. http://dx.doi.org/10.2172/1509420.

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Knox, N. P., J. R. Webb, S. D. Ferguson, L. F. Goins und P. T. Owen. Nuclear facility decommissioning and site remedial actions. Office of Scientific and Technical Information (OSTI), September 1990. http://dx.doi.org/10.2172/6162145.

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Owen, P. T., N. P. Knox, S. D. Ferguson, J. M. Fielden und P. L. Schumann. Nuclear facility decommissioning and site remedial actions. Office of Scientific and Technical Information (OSTI), September 1989. http://dx.doi.org/10.2172/5392482.

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Frazier, R., D. Harrison, R. Meyer, F. Schrag und R. Wilson. Nuclear Materials Development Facility decommissioning final report. Office of Scientific and Technical Information (OSTI), März 1987. http://dx.doi.org/10.2172/6530490.

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Cantor, R. Nuclear reactor decommissioning: an analysis of the regulatory environments. Office of Scientific and Technical Information (OSTI), August 1986. http://dx.doi.org/10.2172/5150096.

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Elder, H. K. Technology, safety and costs of decommissioning reference nuclear fuel cycle facilities. Office of Scientific and Technical Information (OSTI), Mai 1986. http://dx.doi.org/10.2172/5752809.

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Owen, P. T., N. P. Knox, D. C. Michelson und G. S. Turmer. Nuclear facility decommissioning and site remedial actions: A selected bibliography, volume 9. Office of Scientific and Technical Information (OSTI), September 1988. http://dx.doi.org/10.2172/6735525.

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Owen, P. T., J. R. Webb, N. P. Knox, L. F. Goins, R. E. Harrell, P. K. Mallory und C. D. Cravens. Nuclear facility decommissioning and site remedial actions: A selected bibliography, Volume 12. Office of Scientific and Technical Information (OSTI), September 1991. http://dx.doi.org/10.2172/5882373.

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Owen, P. T., D. C. Michelson und N. P. Knox. Nuclear facility decommissioning and site remedial actions. Volume 6. A selected bibliography. Office of Scientific and Technical Information (OSTI), September 1985. http://dx.doi.org/10.2172/5128288.

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