Dissertations / Theses on the topic 'Nuclear decommissioning and dismantling'
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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 systèmeThis 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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Le, Parc Samuel. "Containment during the decommissioning of nuclear power plant: calculation and approach associated." Thesis, KTH, Fysik, 2019. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-266822.
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Yan, Weida. "A Study on Augmented Reality for Supporting Decommissioning Work of Nuclear Power Plants." Kyoto University, 2013. http://hdl.handle.net/2433/179338.
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Kiellman, Tracy Jo. "A health risk assessment for the decommissioning of the Georgia Institute of Technology Research reactor." Thesis, Georgia Institute of Technology, 1998. http://hdl.handle.net/1853/16698.
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Longmire, Pamela. "Nonparametric statistical methods applied to the final status decommissioning survey of Fort St. Vrains prestressed concrete reactor vessel." The Ohio State University, 1998. http://rave.ohiolink.edu/etdc/view?acc_num=osu1407398430.
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Hetherington, Andrew. "Characterisation of reactor graphite to inform strategies for the disposal of reactor decommissioning waste." Thesis, University of Birmingham, 2013. http://etheses.bham.ac.uk//id/eprint/4409/.
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Graphite has been used extensively in UK reactors since the 1950s. The UK nuclear decommissioning programme will result in some 90,000 tonnes of waste graphite being removed from Magnox, AGR, research reactors and plutonium production reactors. It is necessary to understand the radiological characteristics of reactor graphite as a prerequisite for decisions about its interim management as well as final disposition. There is in particular a need to improve confidence in the disposal inventory of the long-lived radionuclides carbon-14 and chlorine-36. Models have been developed to predict the distribution of principal radionuclides for Chapelcross reactor 1 and Wylfa reactor 1, and the calculated inventory compared with published experimental measurements on active samples. The models show good agreement with experimental values for carbon-14 and cobalt-60. However, for the highly mobile and volatile radionuclides chlorine-36 and tritium agreement is poor. The models provide a crude upper limit on the inventory, but certain radionuclides may be released during irradiation. For Wylfa it is predicted that all graphite waste arisings will be ILW. For Chapelcross of the order of 16% of the graphite core may be classified as LLW after the C&M period, but levels of carbon-14 rule out disposal to the LLWR facility.
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SANTOS, IVAN. "Descomissionamento de uma usina de producao de hexafluoreto de uranio." reponame:Repositório Institucional do IPEN, 2008. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11758.
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Made available in DSpace on 2014-10-09T12:55:29Z (GMT). No. of bitstreams: 0Made available in DSpace on 2014-10-09T14:04:54Z (GMT). No. of bitstreams: 0
Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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Metelková, Michaela. "Problematika vytváření rezerv na vyřazování jaderných zařízení." Master's thesis, Vysoká škola ekonomická v Praze, 2008. http://www.nusl.cz/ntk/nusl-11058.
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This work is focused on the issue of creation of reserves to cover the costs of decommissioning of nuclear installations. These costs arise from the need to prevent exposure to radioactivity and other pollutants that adversely affect the environment and human health. The thesis describes various systems of decommissioning of nuclear installations and the creation of reserves in selected countries - the Czech Republic, United Kingdom, Finland and France. The analytical part of the work deals with comparative analysis of the creation of reserves for decommissioning of nuclear facilities in selected countries and their subsequent evaluation. The work is a basis for optimizing the system for establishing reserves for decommissioning of nuclear facilities in the Czech Republic.
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Whitton, John. "Participant perceptions on the nature of stakeholder dialogue carried out by the UK Nuclear Decommissioning Authority (NDA)." Thesis, University of Manchester, 2010. http://www.manchester.ac.uk/escholar/uk-ac-man-scw:213503.
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The engagement of stakeholders in a dialogue on the decommissioning of nuclear facilities and the retrieval and treatment of nuclear waste in the UK has proved challenging. The action orientated research presented here has used a mixed methodological approach to examine participants’ perceptions regarding the nature of dialogue being carried out by the NDA National Stakeholder Group (NSG), with the emancipatory aim of raising participant awareness regarding their role and the nature of the dialogue used. Exploration of the emergent theme of fairness has enabled the researcher to provide a contribution to stakeholder theory. This research adds to the theory of the deliberative institution (Reed, 2008), providing evidence for why the effective influence of stakeholders on decision making, communication about this influence, and the institutionalization of stakeholder participation is as important as the engagement itself. The work also provides an important epistemological contribution regarding the role of dialogue within the concept of social sustainability.
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Braidwood, David Walter. "Remediation and restoration of ocean exposed cliff-top, in the context of Dounreay (Scotland) nuclear power plant decommissioning." Thesis, University of Aberdeen, 2018. http://digitool.abdn.ac.uk:80/webclient/DeliveryManager?pid=237241.
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Restoration ecology continues to become ever more relevant as legislation demands we prevent biodiversity losses. Post-industrial restoration sites pose a number of additional challenges, including balancing ecological need with logistical and financial constraints. In the North of Scotland, one such site is the Dounreay Nuclear Power Plant, now undergoing decommissioning. The intention is to restore cliff-top habitat with native vegetation, blending in with its surroundings and necessitating minimal maintenance. The overall objective of this PhD research was to help develop a plan for the restoration of the site. A key challenge in this particular case was the residual low level radioactivity at depth in some areas, and a restoration layer is required to prevent contamination of bioreceptors at the surface, however, topsoil availability is limited. The cliff top site, and exposure to salt spray driven by strong winds, meant the identification of suitable vegetation communities for different areas could be crucial to its success. Vegetation and soil surveys across nine reference sites along the North coast of Scotland identified five non-peat vegetation communities suitable for Dounreay's restoration. Restoration ecology continues to become ever more relevant as legislation demands we prevent biodiversity losses. Post-industrial restoration sites pose a number of additional challenges, including balancing ecological need with logistical and financial constraints. In the North of Scotland, one such site is the Dounreay Nuclear Power Plant, now undergoing decommissioning. The intention is to restore cliff-top habitat with native vegetation, blending in with its surroundings and necessitating minimal maintenance. The overall objective of this PhD research was to help develop a plan for the restoration of the site. A key challenge in this particular case was the residual low level radioactivity at depth in some areas, and a restoration layer is required to prevent contamination of bioreceptors at the surface, however, topsoil availability is limited. The cliff top site, and exposure to salt spray driven by strong winds, meant the identification of suitable vegetation communities for different areas could be crucial to its success. Vegetation and soil surveys across nine reference sites along the North coast of Scotland identified five non-peat vegetation communities suitable for Dounreay's restoration. This prompted the development of a novel concept: that of utilising restoration sites as 'protorefuges' or 'protorefugia', i.e. restoration sites where threatened species at the leading edge of climate change can be translocated ahead of the climate changing. There, they would be joined by individuals of the wider population naturally dispersed as the climate shifts. Overall, these results enabled the development of a refined restoration plan for Dounreay, which takes into account the particular setting, constraints and timelines involved. With the decommissioning of an increasing number of nuclear sites across Britain and Europe taking place in the coming years, this research should be developed further. In particular our novel concept of protorefugia could even be put into practice, benefiting both restoration and conservation.
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McKillop, Jordan M. "Reducing the activation of the IRIS reactor building using the SCALE/MAVRIC methodology." Thesis, Georgia Institute of Technology, 2009. http://hdl.handle.net/1853/37209.
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The main objective of this research is: (1) to develop a model and perform numerical simulations to evaluate the radiation field and the resulting dose to personnel and activation of materials and structures throughout the IRIS nuclear power plant, and (2) to confirm that the doses are below the regulatory limit, and assess the possibility to reduce the activation of the concrete walls around the reactor vessel to below the free release limit.
IRIS is a new integral pressurized water reactor (PWR) developed by an international team led by Westinghouse with an electrical generation capacity of 335 MWe and passive safety systems. Its design differs from larger loop PWRs in that a single building houses the containment as well as all the associated equipment including the control room that must be staffed continuously. The resulting small footprint has positive safety and economic implications, and the integral layout provides additional shielding and thus the opportunity to significantly reduce the activation, but it also leads to significantly more challenging simulations.
The difficulty in modeling the entire building is the fact that the source is attenuated over 10 orders of magnitude before ever reaching the accessible areas. For an analog Monte Carlo simulation with no acceleration (variance reduction), it would take many processor-years of computation to generate results that are statistically meaningful. Instead, to generate results for this thesis, the Standardized Computer Analyses for Licensing Evaluation (SCALE) with the package Monaco with Automated Variance Reduction using Importance Calculations (MAVRIC) will be used. This package is a hybrid methodology code where the forward and adjoint deterministic calculations provide variance reduction parameters for the Monte Carlo portion to significantly reduce the computational time.
Thus, the first task will be to develop an efficient SCALE/MAVRIC model of the IRIS building. The second task will be to evaluate the dose rate and activation of materials, specifically focusing on activation of concrete walls around the reactor vessel. Finally, results and recommendations will be presented.
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D'Amico, Miriam. "Étude expérimentale et modélisation des explosions hybrides solides/solides : application au cas des mélanges de poussières graphite/métaux." Thesis, Université de Lorraine, 2016. http://www.theses.fr/2016LORR0256/document.
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Dans le cadre des opérations de démantèlement des centrales nucléaires UNGG (Uranium Naturel Graphite Gaz), l’occurrence de phénomènes indésirables, tels que l’inflammation et l’explosion de poudres, ne peut pas être systématiquement exclue. Plus particulièrement, le risque d’inflammation et d’explosion de poussières de graphite, pur ou mélangé avec des impuretés métalliques telles que des particules de magnésium ou de fer, nécessite d’être évalué de façon plus approfondie. Les travaux de cette thèse s’inscrivent donc dans ce contexte et ont deux objectifs principaux : l’évaluation expérimentale de l’explosivité et sa modélisation. 1. L’évaluation expérimentale de l’explosivité des poudres d'intérêt a été réalisée tant en termes de sensibilité à l’inflammation, en couche et en nuage, que de sévérité à l’explosion. En effet, les caractéristiques explosives d’une poussière ou d’un mélange sont fortement influencées par plusieurs paramètres. Ils dépendent d’une part des conditions opératoires, tels que la turbulence, la température et l'énergie d’inflammation, et d’autre part, des propriétés physico-chimiques et de la composition des matériaux. Cette étude s’est focalisée sur des poudres pures de graphite, de magnésium et de fer de taille micrométrique et sur leurs mélanges, dans un éventail de concentrations d’intérêt industriel. Nous avons constaté que l’introduction de métaux peut changer en premier lieu l’étape limitant la vitesse de combustion du graphite. Tout d’abord, les phénomènes cinétiques limitant de l’oxydation du graphite ont été distingués de ceux des métaux (respectivement, réaction hétérogène ou flamme de diffusion gazeuse). En deuxième lieu, il est apparu que la flamme peut être épaissie par la présence du rayonnement lors de la combustion du métal, alors que ce phénomène est négligeable pour le graphite pur. Enfin, la turbulence initiale du nuage de poussière peut être elle aussi modifiée par l'ajout d'une deuxième poudre en vue des caractéristiques granulométriques et de densité différentes. Une étude paramétrique a donc été réalisée afin d'évaluer l'explosibilité des mélanges considérés en prenant en compte les effets de l'humidité relative des poudres, de leur distribution granulométrique, de la puissance de la source d'ignition, de la turbulence initiale du milieu et de la composition. Pour ce faire, nous avons utilisé à la fois des appareils et des technologies conventionnelles, tels que la sphère de 20 litres, la vélocimétrie par images de particules et la thermogravimétrie, mais également des nouvelles installations dédiées à la caractérisation des écoulements turbulents transitoires lors de la dispersion des poudres dans la sphère d'explosion et à l’étude de la propagation d’une flamme en milieu semi-confiné. Il a été clairement démontré que l'ajout de poudres métalliques influence l'aptitude à enflammer le nuage de poussière. L'énergie et la température minimale d'inflammation diminuent fortement lorsque le magnésium est ajouté au graphite ; ce phénomène est moins sensible pour les particules de fer. De plus, la sévérité de l'explosion augmente avec une telle addition. Cet effet de promotion est particulièrement visible sur la cinétique de combustion. 2. La modélisation du phénomène explosif a été réalisée à l’aide de la simulation numérique afin d’estimer une vitesse de propagation de flamme laminaire et d’étudier les effets induits par des facteurs spécifiques d’intérêt industriel, tels que le diamètre des particules ou la concentration en poudre. L’intérêt d’estimer une vitesse de flamme laminaire réside dans son caractère pseudo-intrinsèque. En connaissant les caractéristiques turbulentes d’un milieu industriel complexe, ce paramètre donne la possibilité d’obtenir une vitesse de propagation de flamme turbulente propre au milieu réel et donc d’estimer les effets d’une explosion potentielle. Les résultats expérimentaux ont été utilisés afin de valider le modèle numérique développéDuring the decommissioning operations of the UNGG (Natural Uranium Graphite Gas) nuclear plants, the occurrence of undesirable phenomena, such as dust ignition and explosion, cannot be systematically neglected. In particular, graphite powders, pure or mixed with metals impurities present on the sites, such as magnesium or iron, can represent a potential risk that needs to be further evaluated. This work falls within this context and has two main objectives: the experimental evaluation of the explosion severity and its modeling. 1. The experimental evaluation of the explosivity of such a powders has been carried out both in terms of ignition sensitivity, of dust layer and cloud, and explosion severity. Actually, explosive characteristics of a dust or of a mixture are strongly influenced by several parameters. They depend on the one side on the operating conditions, such as turbulence, temperature and energy of the ignition source, and on the other side, of course, on the materials physicochemical properties and composition. This study focuses on pure micronized powders of graphite, magnesium, and iron and on their mixtures, in a concentration range of industrial interest. It has been demonstrated that the introduction of metals can change, first of all, the rate limiting step of the graphite combustion. Therefore, the kinetic phenomena controlling the graphite oxidation have been distinguished from those of metals (oxygen diffusion or metal vaporization). Secondly, the flame can be thickened by the presence of the radiation during the metal combustion, while this phenomenon is negligible for pure graphite. Finally, the initial turbulence of the dust cloud can be modified by adding a second powder because of the different granulometric characteristics and density. A parametric study was conducted to evaluate the mixtures explosivity taking into account the effects of the relative humidity, the particle size distribution of the powders, the power of the ignition source, the initial turbulence and the composition of the mixture. In order to do this, we used both conventional devices and technologies, such as 20-liters explosion sphere, the particles image velocimetry and the thermogravimetry, but also new facilities dedicated to the characterization of the transient turbulent flow during the dispersion of the powders in the explosion sphere and to study the propagation of a semi-confined flame. It was clearly demonstrated that the addition of metals influences the ability to ignite the dust cloud. The minimum ignition energy and temperature greatly decrease when magnesium powder is added to graphite dust; this phenomenon is less remarkable for iron particles. In addition, the severity of the explosion increases with such an addition. This promotion effect is particularly significant on the combustion kinetics. 2. The modeling of the explosive phenomenon has been performed using numerical simulations in order to estimate a laminar flame propagation velocity and to study the effects induced by specific factors of industrial interest, such as the particle size or the powder concentration. The interest in determining a laminar flame velocity is its pseudo-intrinsic character. Once known the turbulent characteristics of a complex industrial environment, this parameter gives the opportunity to obtain a turbulent flame propagation velocity in a real environment and, therefore, to estimate the effects of a potential explosion. Experimental results were used to validate the numerical model developed during this work
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Pelleterat, de Borde Melchior. "Contribution à la maîtrise du changement organisationnel et de son impact sur la sûreté : le cas de la transition d’une installation nucléaire du fonctionnement vers le démantèlement." Thesis, Paris, ENMP, 2013. http://www.theses.fr/2013ENMP0089.
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Le travail de thèse cherche à dresser une passerelle entre la question du changement organisationnel et celle des risques, à travers l'étude d'une phase complexe du cycle de vie d'une installation nucléaire : le passage du fonctionnement au démantèlement. Ce dernier clôt une dynamique d'exploitation et en hérite les habitudes, les circuits de communication et de décision, mais s'accompagne d'un changement organisationnel conséquent que les acteurs doivent apprendre et intégrer. Il s'agit donc d'interroger la dynamique de transformation organisationnelle en la replaçant dans un mouvement plus global d'évolution d'une installation dans le temps. Nous cherchons à comprendre dans quelle mesure l'organisation mise en place répond ou non aux besoins de l'installation et de ses acteurs. La thèse montre que la transition s'inscrit dans une temporalité propre où les conséquences et la compréhension du changement ne sont pas homogènes. La transition s'accompagne d'une appropriation différée des nouveaux circuits de décision ainsi que d'une recomposition des réseaux d'acteurs. Cette recomposition met en évidence une tendance au contournement temporaire des structures légitimes de l'organisation qui permet la synthèse progressive des nouvelles contraintes et des anciens univers référentiels des acteursThe thesis seeks to establish a bridge between the issue of organizational change and safety, through the study of a complex phase of the life cycle of a nuclear facility: the transition from operation to decommissioning. Decommissioning closes an operating dynamics and inherits habits, communication channels and decision, but is accompanied by an organizational change plant operatives must understand and take possession of. The thesis therefore examines the dynamics of organizational transformation by taking into account a broader movement of evolution of a system over time. We seek to understand how the organization does or does not meet the needs of the facility and its actors. We show that the transition follows in its own temporality and that the consequences and their understanding are not homogeneous. The transition is accompanied by a deferred appropriation by the plant operatives of the new decision circuits, and a radical recomposition of operatives networks. This reconstruction reveals a trend towards a temporary bypass of legitimate organizational structures, as well as a gradual synthesis of the old and new referential universes for the plant operators
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Busse, Alexander Lucas. "Estimativa do inventário de material radioativo para centrais nucleares PWR no descomissionamento." reponame:Repositório Institucional da UFABC, 2016.
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Orientador: Prof. Dr. João Manoel Losada MoreiraDissertação (mestrado) - Universidade Federal do ABC. Programa de Pós-Graduação em Energia, 2016.
Este trabalho faz uma estimativa do inventário de material radioativo oriundo do descomissionamento de reatores nucleares a água pressurizada (PWR). Os volumes e atividades dos resíduos radioativos provenientes do descomissionamento de reatores do tipo PWR semelhantes aos reatores da Central Nuclear Álvaro Alberto - BR foram compilados a partir de dados do reator Trojan. As atividades do vaso, internos e barreira de concreto do reator foram calculados com auxílio dos programas KENO V.a e SERPENT. Os fluxos de nêutrons calculados para os reatores de Angra 2 e Trojan foram utilizados para calcular a ativação do barril, vaso e barreira de concreto dos reatores. Os resultados evidenciaram a contribuição de nêutrons térmicos no vaso do reator devida a reflexão na barreira de concreto. Também foram estimados o volume de elementos combustíveis irradiados para 40 anos de operação de um dos reatores do sítio de Angra. Quase a totalidade da atividade induzida nos sistemas, estruturas e equipamentos do sítio encontra-se no combustível irradiado. Esses, classificados como resíduos de alto nível, totalizam um volume de 591 m3 ou 5,8 % do total de resíduos radioativos incluindo aqueles provenientes do descomissionamento. Os grandes equipamentos do circuito primário representam 2298 m3 ou 22,4% do total de resíduos e contribuem com 99,994 % da atividade dos resíduos de baixo e médio nível. O restante, 71 % do volume ou 7351 m3 são resíduos de nível muito baixo. O espaço total requerido para os resíduos radioativos oriundos do descomissionamento das três usinas nucleares do sítio de Angra seria em torno de 30.000 m3. Esses resíduos requerem armazenamento por aproximadamente 150 anos.
This work estimates the radioactive inventory resulting from the decommissioning process of pressurized water reactors (PWR). The volumes and activities of radioactive waste from the decommissioning of PWR reactors similar to those in the Angra site have been appraised out of data of the Trojan nuclear power plant and correlated to the plant thermal power level. The activities from the reactor vessel, internals and bioshield were estimated with the KENO V.a and SERPENT codes. The neutron fluxes calculated for the Angra 2 and Trojan reactors were used to estimate the activation of the barrel, vessel and bioshield. The total volume of spent fuel elements for 40 years of a 1300 MWPWR was also estimated. Most of activity induced in systems, structures and equipment of the site comes from the spent fuel. The total fuel volume, classified as high-level waste amounted to 591 m3 or 5.8 % of the total radioactive residues including those from the decommissioning. The major equipment of the nuclear steam supply system amounted 2298 m3, or 22.4 % of total waste, and contributed with 99.994 % of the total activity from low and medium level waste. The remaining 71 % of the volume, or 7351 m3, were classified as very low level waste. The total space required for the radioactive waste arising from the decommissioning of the three nuclear power plants of the Angra site was estimated as 30,000 m3. This total waste requires storage for approximately 150 years.
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Silbermann, Gwennaelle. "Effets de la température et de l'irradiation sur le comportement du 14C et de son précurseur 14N dans le graphite nucléaire. Étude de la décontamination thermique du graphite en présence de vapeur d'eau." Thesis, Lyon 1, 2013. http://www.theses.fr/2013LYO10168.
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Le démantèlement des réacteurs Uranium Naturel Graphite Gaz génèrera en France environ 23 000 tonnes de déchets radioactifs graphités. La gestion appropriée de ces déchets nécessite de déterminer leur inventaire radiologique et de disposer de données fiables sur la localisation et la spéciation des radionucléides (RN). Le 14C a été identifié comme RN d'intérêt pour le stockage en raison de son inventaire initial important et du risque de présence d'une fraction organique mobile dans l'environnement, lors de la phase de stockage. A ce titre, l'objectif de cette thèse CIFRE, réalisée en partenariat avec EDF, est de mettre en œuvre des études expérimentales permettant de simuler et d'évaluer l'impact de la température, de l'irradiation et de la corrosion radiolytique du graphite sur le comportement migratoire en réacteur du 14C et de son précurseur azote. Les données ainsi acquises sont intégrées dans la deuxième partie de ce travail consacrée à l'étude d'un procédé de décontamination thermique du graphite en présence de vapeur d'eau. La démarche expérimentale consiste à simuler respectivement la présence de 14C et de 14N par implantation ionique de 13C et d'azote (14N ou 15N) dans un graphite de rondin SLA2 vierge. Cette étude montre que dans la gamme de températures du graphite en réacteur (100 - 500°C) et en absence de corrosion radiolytique, le 13C est stable thermiquement quel que soit l'état de structure du graphite. En revanche, les expériences d'irradiation du graphite chauffé à 500°C au contact d'un gaz représentatif du caloporteur radiolysé montrent le rôle synergique joué par les espèces oxydantes et l'endommagement du graphite favorisant la mobilité du 13C par gazéification des surfaces et/ou oxydation sélective du 13C plus faiblement lié. En ce qui concerne l'azote constitutif, il a tout d'abord été démontré que sa concentration en surface atteint plusieurs centaines de ppm (< 500 ppm at.) et décroît en profondeur jusqu'à environ 160 ppm at.. Contrairement au 13C implanté, l'azote implanté migre à 500°C lorsque le graphite est fortement déstructuré (environ 8 dpa) alors qu'il reste stable pour un taux de déstructuration moindre (0,14 dpa). Les expériences montrent également le rôle synergique des excitations électroniques et de la température qui accélèrent le transport de l'azote vers la surface du graphite. Cette migration de l'azote semble se faire sous forme moléculaire d'espèces C-N, C=N voire C N. Après huit heures d'irradiation ces espèces ne sont toutefois pas ou peu relâchées et restent bloquées à la surface. L'étude du procédé de décontamination thermique en présence de vapeur d'eau a nécessité la mise en place d'un dispositif de thermogravimétrie couplé à un générateur de vapeur d'eau ainsi que l'optimisation des paramètres de l'étude. Les influences de la température (700°C et 900°C) et de l'humidité relative (50 % HR et 90 % HR) ont été testées à un débit de gaz humide fixe de 50 mL/min. Dans ces conditions, l'oxydation sélective du carbone implanté a été confirméeThe dismantling of UNGG reactors in France will generate about 23 000 tons of radioactive graphite wastes. To manage these wastes, the radiological inventory and data on radionuclides (RN) location and speciation should be determined. 14C was identified as an important RN for disposal due to its high initial activity and the risk of release of a mobile organic fraction in environment, after water ingress into the disposal. Hence, the objective of this thesis, carried out in partnership with EDF, is to implement experimental studies to simulate and evaluate the impact of temperature, irradiation and graphite radiolytic corrosion on the in reactor behavior of 14C and its precursor, 14N. The obtained data are then used to study the thermal decontamination of graphite in presence of water vapor. The experimental approach aims at simulating the presence of 14C and 14N by the respective ion implantation of 13C and 14N or 15N in virgin graphite. This study shows that, in the temperature range reached during reactor operation, (100-500°C) and without radiolytic corrosion, 13C is thermally stable whatever the initial graphite structure. Moreover, irradiation experiments were performed on heated graphite (500°C) put in contact with a gas representative of the radiolysed coolant gas. They show the synergistic role played by the oxidative species and the graphite structure disorder on the enhancement of 13C mobility resulting in the gasification of the graphite surface and/or the selective oxidation of 13C more weakly bound than 12C. Concerning the pristine nitrogen, we showed first that the surface concentration reaches several hundred ppm (<500 ppm at) and decreases at deeper depths to about 160 ppm at.. Unlike implanted 13C, implanted nitrogen migrates at 500 ° C when the graphite is highly disordered (about 8 dpa) while remaining stable for a lower disorder rate (0.14 dpa). Experiments also show the synergistic role by electronic excitations and temperature that accelerate the transport of nitrogen to the surface of the graphite. Nitrogen seems to migrate in the form of molecular species (CN, C = N or C N). After eight hours of irradiation these species are, however, little or not released and blocked at the surface. The study of the thermal decontamination of graphite in presence of water vapor was performed with a thermogravimetric device coupled to a steam water generator device. The influence of temperature (700 ° C and 900 ° C) and of the relative humidity (50% RH and 90% RH) was tested with a wet gas fixed flow rate of 50 ml/min. Under these conditions, the selective oxidation of implanted carbon was confirmed
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AlAli, May. "A Comparative Analysis of Decommissioning Scenarios Based on Radiation Dose Modeling and Multi-criteria Decision Analysis for Oskarshamn Nuclear Reactor 3 : Lessons Learned from Operating Experience in the Reuse of Decommissioned Sites." Thesis, Uppsala universitet, Institutionen för elektroteknik, 2020. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-438560.
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An increasing number of decommissioning activities are being undertaken worldwide at facilities where radioactive material has been produced, used, or managed, which will also be the case of the Oskarshamn nuclear reactor O3 in Sweden, the remaining operational unit at that power plant. Decommissioning, which seems to be the sign of ending, is, in fact, a future-oriented process that can prove successful the safe management of nuclear facilities throughout their entire lifetimes but especially after shutdown. In this context, this research aims to develop a comparative profile of decommissioning scenarios based on a radiation dose modeling approach using RESRAD-BUILD software. Conducting then a multi-criteria decision analysis (MCDA) of scenario comparison for the decommissioning of O3 is essential for updating on-going decommissioning projects, making strategic decisions for future ones, and bringing in sustainable site reuse options. The exposure-to-dose model will be used to evaluate and optimize safety parameters, site release levels particularly, related to personnel and the environment. Testing multiple sets of parameters in the decommissioning plan will be used to compare results and to assess the sensitivity of the strategy to variable inputs. This comparison with the MCDA Analytic Hierarchy Process model (AHP) results will allow the identification of the most optimal reuse scenario for Oskarshamn unit 3 after its useful life ends. Also, the lessons learned from operating experience in the reuse of decommissioned nuclear sites can be incorporated systematically into the eventual decommissioning of O3, and which, according to current planning will run until 2035 or 2045.
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Macario, Barros Andréa. "Modular device for automated and reliable mapping of indoor installations." Electronic Thesis or Diss., université Paris-Saclay, 2023. http://www.theses.fr/2023UPASP186.
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Dans le cadre des opérations de démantèlement, l'établissement d'une cartographie radiologique précise est essentiel, car cela facilite l'identification des zones présentant des niveaux de rayonnement élevés. Cette tâche repose principalement sur les opérateurs de radioprotection qui quadrillent les zones à contrôler et réalisent des mesures aux positions correspondantes. Néanmoins, cette méthode manuelle est sujette à des erreurs humaines, peut être physiquement épuisante pour les opérateurs et les exposer à des environnements potentiellement contaminés. Par conséquent, la littérature explore progressivement des approches alternatives, comme l'intégration de mesures radiologiques avec des techniques de localisation et de cartographie simultanées (SLAM). Ces techniques offrent la capacité de cartographier l'environnement et de déterminer l'emplacement d'un capteur sans recourir aux systèmes GPS ou GNSS, généralement inopérants dans les installations nucléaires intérieures. Cependant, les solutions incorporant le SLAM dans la littérature sont souvent limitées à la localisation d'un seul type de mesure nucléaire, présentent un encombrement conséquent (dimensions et poids), et nécessitent une étape de post-traitement. En réponse à ces limitations, cette thèse propose le développement d'un dispositif modulaire pour la reconstruction et localisation 3D de mesures radiologiques, avec la volonté d'identifier l'algorithme SLAM le plus adapté au cadre du démantèlement embarqué. Sachant que la littérature présente un large éventail d'algorithmes SLAM et d'odométrie, sélectionner le plus fiable pour la reconstruction des installations nucléaires n'est pas chose aisée. Pour répondre à cette question, une révision de la littérature des algorithmes visuel-SLAM a été menée. Par la suite, ces algorithmes ont été évalués quant à leur résilience dans les conditions rencontrées dans les environnements en démantèlement. Cinq algorithmes ont été retenus pour être mis en œuvre parmi la gamme d'algorithmes identifiés, chacun ayant le potentiel de produire des performances satisfaisantes dans le contexte de cette thèse. Les algorithmes sélectionnés comprennent les Direct Sparse Odometry (DSO), Visual-Inertial Direct Sparse Odometry (VI-DSO), Large Scale Direct Monocular SLAM (LSD-SLAM), Semi-direct Visual Odometry (SVO), and Visual Inertial Semi-direct Visual Odometry (VI-SVO). Pour permettre leur comparaison, un nouvel ensemble de données représentant les caractéristiques des opérations de cartographie radiologique a été conçu. Cet ensemble de données a été créé à partir d'un nouveau prototype intégrant des images stéréo et des données inertielles, sphériques et radiologiques. Il a permis l'analyse comparative des algorithmes en tenant compte de leur précision de localisation et cartographie, de leur embarquabilité et de leur capacité à localiser les points chauds. Le VI-SVO a présenté les erreurs moyennes les plus faibles pour la localisation et une performance équivalente aux autres algorithmes pour la cartographie. Le VI-DSO s'est avéré être l'algorithme le plus approprié pour une implémentation embarquée. Contrairement au VI-SVO, cet algorithme n'a pas pu traiter toutes les séquences de l'ensemble des données. Parmi les algorithmes évalués, le VI-SVO a été le seul à traiter avec succès toutes les séquences et à localiser les zones de contamination. Il est donc l'algorithme le plus approprié pour la cartographie radiologiqueIn the context of the Dismantling and Decommissioning processes, establishing a precise radiological mapping is essential, as it facilitates the identification of areas with elevated radiation levels. This task predominantly relies on manual procedures performed by radiation protection operators, who construct matrices and allocate the measurements to their corresponding spatial positions. Nonetheless, this conventional method is more susceptible to human errors, can be physically exhausting for the operators, and expose them to potentially contaminated environments. Consequently, the literature is progressively exploring alternative approaches, as the integration of radiological measurements with Simultaneous Localization and Mapping (SLAM) techniques. SLAM technology offers the capability to concurrently mapping the surrounding environment and determining the location of a sensor, without the reliance on GPS or GNSS systems, which are typically non-functional within indoor nuclear facilities. Nevertheless, existing SLAM solutions in the literature are often limited to the localization of one type of nuclear measurement, tend to be cumbersome in design, and require post-processing procedures. In response to these limitations, this thesis proposes the development of a modular device for online 3D environment reconstruction and radioactivity measurement localization, focusing on the identification of the most appropriate SLAM algorithm in the context of embedded nuclear dismantling. While the literature presents an array of SLAM and odometry algorithms, selecting the most robust one for reconstructing nuclear facilities is not straightforward. To address this concern, a comprehensive review of state-of-the-art visual-sensor-based SLAM algorithms was conducted. Subsequently, these algorithms were critically evaluated concerning their resilience in the specific conditions encountered in dismantling environments. Five were chosen for implementation from the array of identified algorithms, each with the potential to yield satisfactory performance in the context of nuclear facility reconstruction. These selected algorithms include Direct Sparse Odometry (DSO), Visual-Inertial Direct Sparse Odometry (VI-DSO), Large Scale Direct Monocular SLAM (LSD-SLAM), Semi-direct Visual Odometry (SVO), and Visual Inertial Semi-direct Visual Odometry (VI-SVO). A novel dataset was conceived in the frame of this thesis to facilitate a comparative assessment. This dataset aims to accurately represent the characteristics of radiological mapping operations within nuclear facilities. This dataset was conceived using a new handheld prototype integrating stereo images and inertial, spherical, and radiological data. The proposed dataset allowed the benchmarking of the algorithms considering algorithms' tracking and mapping accuracies, embeddability, and ability to locate hotspots. The VI-SVO presented the lowest average errors for the tracking and an equivalent performance as the other algorithms for the mapping. The VI-DSO has been demonstrated to be the most suitable algorithm for an embedded implementation. However, unlike the VI-SVO, this algorithm could not process all the real-case sequences. Among the evaluated algorithms, the VI-SVO was the only one to successfully process all the sequences in the dataset and localize the contamination data, being the most suitable algorithm for the radiological mapping
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Le, Guillou Maël. "Migration du deutérium dans le graphite nucléaire : conséquences sur le comportement du tritium en réacteur UNGG et sur la décontamination des graphites irradiés." Thesis, Lyon 1, 2014. http://www.theses.fr/2014LYO10227/document.
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En France, 23 000 tonnes de graphites irradiés générés par le démantèlement des réacteurs nucléaires de première génération Uranium Naturel-Graphite-Gaz (UNGG) sont en attente d'une solution de gestion à long terme. Cette thèse porte sur le comportement du tritium, l'un des principaux contributeurs à l'inventaire radiologique des graphites à l'arrêt des réacteurs. Afin d'anticiper des rejets de tritium lors du démantèlement ou de la gestion des déchets, il est indispensable d'obtenir des données sur sa migration, sa localisation et son inventaire. Notre étude repose sur la simulation du tritium par implantation de l'ordre de 3 % at. de deutérium jusqu'à environ 3 μm dans un graphite nucléaire vierge. Celui-ci a ensuite subi des recuits jusqu'à 300 h et 1300 ° C sous atmosphère inerte, gaz caloporteur UNGG et gaz humide, dans le but de reproduire des conditions proches de celles rencontrées en réacteur et lors des opérations de gestion des déchets. Les profils et la répartition spatiale du deutérium ont été analysés via la réaction nucléaire 2H(3He,p)4He. Les principaux résultats montrent un relâchement thermique du deutérium se produisant selon trois régimes contrôlés par le dépiégeage de sites superficiels ou interstitiels. L'extrapolation des données au cas du tritium tend à montrer que son relâchement thermique en réacteur pourrait avoir été inférieur à 30 % et localisé à proximité des surfaces libres du graphite. L'essentiel de l'inventaire en tritium à l'arrêt des réacteurs serait retenu en profondeur dans les graphites irradiés, dont la décontamination nécessiterait alors des températures supérieures à 1300 °C, et serait plus efficace sous gaz inerte que sous gaz humideIn France, 23 000 t of irradiated graphite that will be generated by the decommissioning of the first generation Uranium Naturel-Graphite-Gaz (UNGG) nuclear reactors are waiting for a long term management solution. This work focuses on the behavior of tritium, which is one of the main contributors to the radiological inventory of graphite waste after reactor shutdown. In order to anticipate tritium release during dismantling or waste management, it is mandatory to collect data on its migration, location and inventory. Our study is based on the simulation of tritium by implantation of approximately 3 at. % of deuterium up to around 3 μm in a virgin nuclear graphite. This material was then annealed up to 300 h and 1300 °C in inert atmosphere, UNGG coolant gas and humid gas, aiming to reproduce thermal conditions close to those encountered in reactor and during waste management operations. The deuterium profiles and spatial distribution were analyzed using the nuclear reaction 2H(3He,p)4He. The main results evidence a thermal release of implanted deuterium occurring essentially through three regimes controlled by the detrapping of atomic deuterium located in superficial or interstitial sites. The extrapolation of our data to tritium suggests that its purely thermal release during reactor operations may have been lower than 30 % and would be located close to the graphite free surfaces. Consequently, most of the tritium inventory after reactor shutdown could be trapped deeply within the irradiated graphite structure. Decontamination of graphite waste should then require temperatures higher than 1300°C, and would be more efficient in dry inert gas than in humid gas
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Noonan, Christine F. "Federal City revisited : atomic energy and community identity in Richland, Washington." Virtual Press, 2000. http://liblink.bsu.edu/uhtbin/catkey/1180787.
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This study examines the relationship between atomic energy production and community identity in Richland, Washington. Over the past fifty years, the identity of southeastern Washington has been intimately tied to production and industry at the Hanford Site. Today, however, environmental restoration and waste management programs have replaced plutonium production. The decline of the nuclear industry has influenced reinterpretations of local history and community identity through public display, commodity goods, and the re-scripting of historical texts.Department of Anthropology
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Harthan, Ralph Oliver. "Integration of Renewable Energies into the German Power System and Their Influence on Investments in New Power Plants." Doctoral thesis, Universitätsbibliothek Leipzig, 2015. http://nbn-resolving.de/urn:nbn:de:bsz:15-qucosa-160117.
Full textAbstract:
The increasing share of renewable energies in the power sector influences the economic viability of investments in new conventional power plants. Many studies have investigated these issues by considering power plant operation or the long-term development of the power plant fleet. However, power plant decommissioning, investment and operation are intrinsically linked. This doctoral thesis therefore presents a modelling framework for an integrated consideration of power plant decommissioning, investment and operation. In a case study focusing on Germany, the effects of the integration of renewable energies on power plant decommissioning, investment and operation are evaluated in the context of different assumptions regarding the remaining lifetime of nuclear power plants. With regard to the use of nuclear power, a phase-out scenario and a scenario with lifetime extension of nuclear power plants (by on average 12 years) are considered. The results show that static decommissioning (i.e. considering fixed technical lifetimes) underestimates the capacity available in the power sector in the scenario without lifetime extension since retrofit measures (versus decommissioning) are not taken into account. In contrast, capacity available in the case of nuclear lifetime extension is overestimated since mothballing (versus regular operation) is not considered. If the impact on decommissioning decisions of profit margins accrued during power plant operation are considered (“dynamic decommissioning”), the electricity price reduction effect due to a lifetime extension is reduced by more than half in comparison to static decommissioning. Scarcity situations do not differ significantly between the scenarios with and without lifetime extension with dynamic decommissioning; in contrast, there is a significantly higher need for imports without lifetime extension with static decommissioning. The case study demonstrates that further system flexibility is needed for the integration of renewable energies. It can be further concluded that the share of flexible power plants is higher with the phase-out of nuclear power plants. With regard to the decommissioning dynamics, the phase-out can be considered as beneficial for the economic viability of fossil power plants. Furthermore, the phase-out does not, overall, lead to environmental disadvantages in the medium term, but may be beneficial in the long run since lock-in effects are avoided. Further research is required with regard to the consideration of future flexibility options and a new market designDer steigende Anteil erneuerbarer Energien beeinflusst die Wirtschaftlichkeit von Investitionen in neue konventionelle Kraftwerke. Zahlreiche Studien haben diese Aspekte in Bezug auf den Kraftwerksbetrieb oder die langfristige Entwicklung des Kraftwerksparks untersucht. Stilllegungen, Investitionen und Betrieb im Kraftwerkspark bedingen jedoch einander. Aus diesem Grund wird in dieser Doktorarbeit ein Modellierungsansatz für eine integrierte Betrachtung von Kraftwerksstilllegung, -investition und -betrieb vorgestellt. In einer Fallstudie für Deutschland werden die Auswirkungen einer Integration erneuerbarer Energien auf Kraftwerksstilllegung, -investition und -betrieb im Zusammenhang mit unterschiedlichen Annahmen über die Restlaufzeit von Kernkraftwerken untersucht. Bezogen auf die Nutzung der Kernenergie wird hierbei ein Ausstiegsszenario sowie ein Laufzeitverlängerungsszenario (Verlän-gerung der Laufzeit um durchschnittlich 12 Jahre) betrachtet. Die Ergebnisse zeigen, dass die statische Stilllegung (d.h. die Betrachtung fester technischer Lebensdauern) im Fall eines Verzichts auf die Laufzeitverlängerung die im Kraftwerkspark verfügbare Leistung unterschätzt, da Retrofit-Maßnahmen (im Vergleich zur Stilllegung) nicht berücksichtigt werden. Die verfügbare Leistung im Falle einer Laufzeitverlängerung wird dagegen überschätzt, da die Möglichkeit der Kaltreserve (im Vergleich zum regulären Betrieb) vernachlässigt wird. Werden die Rückwirkungen der im Betrieb erwirtschaftbaren Deckungsbeiträge auf Stilllegungsentscheidungen (“dynamische Stilllegung”) betrachtet, so wird der strompreissenkende Effekt durch die Laufzeitverlängerung im Vergleich zur statischen Stilllegung mehr als halbiert. Knappheitssitutationen unterscheiden sich nicht wesentlich mit und ohne Laufzeitverlängerung im Fall der dynamischen Stilllegung, während bei statischer Stilllegung ohne Laufzeitzeitverlängerung ein deutlich größerer Importbedarf besteht. Die Fallstudie zeigt, dass weitere Systemflexibilitäten für die Integration erneuerbarer Energien benötigt werden. Der Anteil flexibler Kraftwerke ist größer im Fall des Kernenergieausstiegs. Der Kernenergieausstieg wirkt sich in Bezug auf die Stilllegungsdynamik positiv auf die Wirtschaftlichkeit fossiler Kraftwerke aus. Insgesamt führt der Kernenergieausstieg zu keinen mittelfristig nachteiligen Umwelteffekten, er kann sich jedoch langfristig positiv auswirken, da Lock-in-Effekte vermieden werden. Es besteht weiterer Forschungsbedarf in Bezug auf die Berücksichtigung künftiger Flexibilitätsoptionen und ein neues Marktdesign
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Panza, Fabien. "Développement de la spectrométrie gamma in situ pour la cartographie de site." Phd thesis, Université de Strasbourg, 2012. http://tel.archives-ouvertes.fr/tel-00975929.
Full textAbstract:
La spectrométrie gamma à haute résolution offre un outil d'analyse performant pour effectuer des mesurages environnementaux. Dans le cadre de la caractérisation radiologique d'un site (naturelle ou artificielle) ainsi que pour le démantèlement d'installations nucléaires, la cartographie des radionucléides est un atout important. Le principe consiste à déplacer un spectromètre HPGe sur le site à étudier et, à partir des données nucléaires et de positionnements, d'identifier, de localiser et de quantifier les radionucléides présents dans le sol. Le développement de cet outil fait suite à une intercomparaison où un exercice orienté intervention a montré les limites des outils actuels. Une partie de ce travail s'est portée sur la représentation cartographique des données nucléaires. La connaissance des paramètres d'un spectre in situ a permis la création d'un simulateur modélisant la réponse d'un spectromètre se déplaçant au-dessus d'un sol contaminé. Ce simulateur a lui-même permis de développer les algorithmes de cartographie et de les tester dans des situations extrêmes et non réalisables. Ainsi, ce travail ouvre sur la réalisation d'un prototype viable donnant en temps réel les informations nécessaires sur l'identité et la position possible des radionucléides. La recherche réalisée sur la déconvolution des données permet de rendre en post traitement une carte de l'activité du sol par radionucléide mais également une indication sur la profondeur de la source. Le prototype nommé OSCAR (Outil Spectrométrique de Cartographie de Radionucléides) a ainsi été testé sur des sites contaminés (Suisse et Japon) et les résultats obtenus sont en accord avec des mesures de référence.
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Chen, Yeh-Cheng, and 陳彥誠. "Intelligent Radioactive Waste Process System for Nuclear Power Plant Decommissioning." Thesis, 2014. http://ndltd.ncl.edu.tw/handle/a485d8.
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Abelquist, Eric Warner. "Dose Modeling and Statistical Assessment of Hot Spots for Decommissioning Applications." 2008. http://trace.tennessee.edu/utk_graddiss/409.
Full textAbstract:
A primary goal of this research was to develop a technically defensible approach for modeling the receptor dose due to smaller "hot spots" of residual radioactivity. Nearly 700 combinations of environmental pathways, radionuclides and hot spot sizes were evaluated in this work. The hot spot sizes studied ranged from 0.01 m2 to 10 m2, and included both building and land area exposure pathways. Dose modeling codes RESRAD, RESRAD-BUILD, and MicroShield were used to assess hot spot doses and develop pathway-specific area factors for eleven radionuclides. These area factors are proposed for use within the existing Multiagency Radiation Survey and Site Investigation Manual (MARSSIM) context of final status survey design and implementation. The research identified pathways that are particularly "hot spot sensitive"—i.e., particularly sensitive to changes in the areal size of the contaminated area. The external radiation pathway was the most hot spot sensitive for eight of the eleven radionuclides studied. These area factors were evaluated both when the receptor was located directly on the soil hot spot and ranged from 6.6 to 11.4 for 1 m2 hot spot; and ranged from 650 to 785 when the receptor was located 6 m from the 1 m2 hot spot. The external radiation pathway was also the most sensitive of the building occupancy pathways. For the smallest building hot spot studied (100 cm2), the area factors were approximately 1100 for each of the radionuclides. A Bayesian statistical approach for assessing the acceptability of hot spots is proposed. A posterior distribution is generated based on the final status survey data that provides an estimate of the 99th percentile of the contaminant distribution. Hot spot compliance is demonstrated by comparing the upper tolerance limit——defined as the 95% upper confidence level on the 99th percentile of the contaminant distribution in the survey unit—with the DCGL99th value. The DCGL99th is the hot spot dose limit developed using the dose modeling research to establish area factors mentioned above. The proposed approach provides a hot spot assessment approach that considers hot spots that may be present, but not found. Examples are provided to illustrate this approach.
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Hu, Yu-Ching, and 胡毓青. "The Evaluation of Filtering Decommissioning Strategies for Domestic Nuclear Power Plant." Thesis, 2008. http://ndltd.ncl.edu.tw/handle/61976731569763258968.
Full textAbstract:
碩士中原大學
企業管理研究所
96
The first nuclear power plant in Taiwan has been operating for 30 years since 1978. Some facilities and components are worn out and requires constant maintenance. A nuclear power plant has its own life span, but it will be shut down, dismantled and decommissioned, when the time comes. According to the life span of 40 years operating time, Chinshan Nuclear Power Plant is scheduled to be shut down in 2018. Maanshan and Kuosheng Nuclear Power Plants are scheduled to be shut down separately in 2020 and 2023.Taiwan has limited land resource and limited experiences in decommissioning. Therefore, to carry out decommissioning of nuclear power plant successively and effectively as planed is our top priority. In order to achieve our plan, it is important for us to set up a realistic evaluate structure to filtering through different decommissioning approaches for making best choices possible. The main purpose of this research is to construct an suitable level structural evaluation system for choosing best decommissioning approach for domestic nuclear power plants. The research is proceeded through interviewing experts to find out optimum ways of decommissioning as well as the key structure and criterion on this matter. The structure face in this research provides an 360 degrees managerial views for best results of reducing risks and making the best possible choice in nuclear power plant decommission in Taiwan. The level structural evaluation system this research constructs includes four evaluation faces-- "Safety", "Technology", "Environment" and "Cost"; and three substitutive plans-- "Immediate dismantling", "Defered dismantling" and "Entombment". The evaluation team in this research is formed by sum of 53 experts from Institute of Nuclear Energy Research and Taipower company. Their evaluation results from the interviews conducted by this research were ran by fuzzy for further evaluation to find out the most suitable decommissioning approaches for us. The important results of this study are as follows : 1. The evaluation team think the most important evaluation face among the four is "Safety", followed by "Environment", "Technology", and "Cost". This shows the safety concern when decommissioning a nuclear power plant is essential. 2. Overall, the best decommissioning strategy would be "Immediate dismantling", followed by "Defered dismantling", then "Entombment". The reasons for "Immediate dismantling" to be the best choice are politically correct and the quickest environmental recovery for further use.
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Chu, Ching-Hau, and 朱璟豪. "Evaluate the feasibility of CSDSS application for nuclear facility decommissioning through CFD." Thesis, 2017. http://ndltd.ncl.edu.tw/handle/6he82k.
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Yeh, Chin-Hsien, and 葉璟賢. "Identification of Vital Areas During Nuclear Power Plants Transitional Stage of Decommissioning." Thesis, 2018. http://ndltd.ncl.edu.tw/handle/ky2x2v.
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Ou, Yan-Zhen, and 歐彥蓁. "A Study of Packing Problems for the Decommissioning Nuclear Power Plant Unit." Thesis, 2018. http://ndltd.ncl.edu.tw/handle/42v9q4.
Full textAbstract:
碩士國立臺北科技大學
經營管理系碩士班
106
This study mainly investigates the three-dimensional packaging problem for the decommissioning nuclear power plant unit. The constraints include active conditions and weight limitations of the given containers. The objective is to minimize the number of given containers to reduce the storage space and costs. To ensure the safety and smoothness of decommissioning work of nuclear facilities and to find out the optimal spatial configuration, this study explores the heuristic algorithms and deterministic algorithms for placing dismantling units to improve space utilization rate. The proposed method can effectively solve the problem to reduce the packing cost. Several numerical examples are presented to illustrate the effectiveness and efficiency of the proposed method.
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Patre, Monika [Verfasser]. "Dismantling of the components of the nuclear pore comlex during apoptosis / vorgelegt von Monika Patre." 2005. http://d-nb.info/977868206/34.
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Liu, Yen-Cheng, and 劉彥呈. "Simulation and Verification Methodology of RESRAD Computer Code for Residual Radioactive Contamination of Decommissioning Nuclear Power Plant." Thesis, 2018. http://ndltd.ncl.edu.tw/handle/s67grb.
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ZHUANG, KAI-YU, and 莊凱宇. "The Research of Risk Assessment of Critical Infrastructure： The Case of Nuclear Power Plant Decommissioning Risk Control." Thesis, 2019. http://ndltd.ncl.edu.tw/handle/enz7e6.
Full textAbstract:
碩士中央警察大學
警察政策研究所
107
Critical Infrastructure (CI) is an important asset of the country and an infrastructure to maintain stable operation of the country. If critical infrastructure breaks out or encounters various disasters, it will not only leave inconvenience to the public's life, but also cause panic among the people's mind. To make matters worse, affect the economic development and security of the entire country, such as the 911 terrorist attacks incident in the United States in 2001. The 2004 tsunami in South Asia and Japan’s 311 earthquake in 2011 were caused by different factors that made the country's critical infrastructure to be inoperable. The losses and impacts triggered by this were very wide and deep, and the reconstruction after the disaster took a lot of effort.It is necessary to invest a lot of resources to reduce the impact of disasters. Nuclear energy is a project that is highly valued in Taiwan's key infrastructure. In particular, the No. 1 machine license of the No. 1 Plant expired on December 5, 2018, and the No. 2 license expires on July 15, 2019 is facing decommissioning phased tasks. This study aims to establish not only a complete and objective and effective risk indicator but also network structure for the risk management as well as control factors of nuclear power plants during decommissioning. What’s more, a questionnaire based on the Consistent Fuzzy Preference Relationships Method (CFPR) is designed. The personnel in charge of the actual decommissioning of nuclear power plants including experts and scholars in related fields who use the multi-criteria decision-making analysis software to complete the construction of the weight system, so as to understand the factor performance value of the nuclear power plant during the period of decommissioning. According to the results of the CFPR research method, establish various indicators of risk management and control factors during the decommissioning of nuclear power plants, and make the following recommendations: 1. Ensure the stability of the cooling system during the decommissioning of nuclear power plants. 2. Strengthen the protection of workplaces during the decommissioning process, avoid the occurrence of radiation accidents, and ensure the health of the people. 3. Pay attention to the physical protection of nuclear power plants, and strictly prevent external personnel from intruding the factory to destroy or steal. 4. Enhance the safety concept of personnel in nuclear power plant areas to ensure the ability to respond in the event of an emergency. 5. Pay attention to the difference between the current status of decommissioning of nuclear power plants in China and foreign countries. It is expected that the results of this research will provide government units of nuclear power plants as a reference for decommissioning, so as to control the risk factors and minimize the impact.
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Findley, Michael. "Group differences in safety climate among workers in the nuclear decommissioning and demolition industry in the United States." 2004. http://etd.utk.edu/2004/FindleyMichael.pdf.
Full textAbstract:
Thesis (Ph. D.)--University of Tennessee, Knoxville, 2004.Title from title page screen (viewed Sep. 28, 2004). Thesis advisor: Susan M. Smith. Document formatted into pages (xiv, 228 p. : ill.). Vita. Includes bibliographical references (p. 128-138).
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HUN-GHY, WU, and 巫鴻志. "Apply Delphi Method and Boyd Cycle Theory on Communication Strategy of Decommissioning Nuclear Power Plants in Taiwan -A Case Study of Chinshan Nuclear Power Plant." Thesis, 2017. http://ndltd.ncl.edu.tw/handle/m8k773.
Full textAbstract:
碩士國立臺北科技大學
管理學院工業工程與管理EMBA專班
105
According to the current government stability reduction policy, to ensure (1) unlimited power;(2) to maintain a reasonable price; (3) to achieve international commitments to reduce carbon and other three principles without fear, Chinshan Nuclear Power Plant will be decommissioned after the date of service expiry in December 107 and July 108. Taipower company must submit a decommissioning plan in period of three years, the decommissioning plan must be reviewed and approved before nuclear power plant decommissioned. Although there are many reference cases in foreign countries, but in the different countries of the laws and regulations, social and theory, customs and geographical environment under different conditions, whether 100% reference to foreign experience and then transplanted to Domestic implementation, still can not confirm. There are still many internal (internal horizontal units, employees, etc.) and the external environment (in charge of a few off, local people, representatives of public opinion, etc.) and other personnel and units to be communicated. Thus, the decommissioning plan can be carried out on time According to the theory of OODA (Observe, Orient, Decide, Act), this study is based on the observation of (Observe) and orientation (Orient ), With the experts questionnaires, through the academic community, research institutions, industry, nuclear professionals and local experts and scholars or gentry, with its different point of view, organize and summarize its intention, and then by Spearman correlation analysis, analysis of its relevance and significance, and then explore the different stakeholders of the communication issues to provide a combination of decision-making units.
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Pillittere, Joseph T. "An assessment of citizen action committees as a risk communication strategy in the decommissioning of Connecticut Yankee nuclear power plant /." 2002. http://www.consuls.org/record=b2585639.
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Thesis (M.S.)--Central Connecticut State University, 2002.Thesis advisor: Robert Fischbach. " ... in partial fulfillment of the requirements for the degree of Master of Science in Organizational Communications." Includes bibliographical references (leaves 86-90). Also available via the World Wide Web.
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Myslivcová, Kateřina. "Vliv odstavení jaderných elektráren na energetickou bezpečnost: komparace Německa a ČR." Master's thesis, 2019. http://www.nusl.cz/ntk/nusl-396778.
Full textAbstract:
The diploma thesis examines the effect of decommissioning of nuclear power plants on the energy security of two European countries - Germany and the Czech Republic. First, the author focuses on theoretical anchoring of energy security. The practical part then examines both countries from the perspective of their energy mixes and policies and infrastructure interconnection. This is the starting point for the comparison of the effects of the decommissioning of nuclear power plants. Second, the author presents three scenarios to replace nuclear energy, exploring how German and Czech energy security will change. The first scenario is the use of renewable energy only. The second option is a combination of renewable energy and coal. Finally, the third option is again a certain proportion of renewable energy along with natural gas. Last but not least, the author concludes in the presented paper that, despite the various possible combinations of how to replace nuclear energy, both countries would be worse off within the framework of energy security.
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Harthan, Ralph Oliver. "Integration of Renewable Energies into the German Power System and Their Influence on Investments in New Power Plants: Integrated Consideration of Effects on Power Plant Investment and Operation." Doctoral thesis, 2014. https://ul.qucosa.de/id/qucosa%3A13132.
Full textAbstract:
The increasing share of renewable energies in the power sector influences the economic viability of investments in new conventional power plants. Many studies have investigated these issues by considering power plant operation or the long-term development of the power plant fleet. However, power plant decommissioning, investment and operation are intrinsically linked. This doctoral thesis therefore presents a modelling framework for an integrated consideration of power plant decommissioning, investment and operation.
In a case study focusing on Germany, the effects of the integration of renewable energies on power plant decommissioning, investment and operation are evaluated in the context of different assumptions regarding the remaining lifetime of nuclear power plants. With regard to the use of nuclear power, a phase-out scenario and a scenario with lifetime extension of nuclear power plants (by on average 12 years) are considered. The results show that static decommissioning (i.e. considering fixed technical lifetimes) underestimates the capacity available in the power sector in the scenario without lifetime extension since retrofit measures (versus decommissioning) are not taken into account. In contrast, capacity available in the case of nuclear lifetime extension is overestimated since mothballing (versus regular operation) is not considered. If the impact on decommissioning decisions of profit margins accrued during power plant operation are considered (“dynamic decommissioning”), the electricity price reduction effect due to a lifetime extension is reduced by more than half in comparison to static decommissioning. Scarcity situations do not differ significantly between the scenarios with and without lifetime extension with dynamic decommissioning; in contrast, there is a significantly higher need for imports without lifetime extension with static decommissioning.
The case study demonstrates that further system flexibility is needed for the integration of renewable energies. It can be further concluded that the share of flexible power plants is higher with the phase-out of nuclear power plants. With regard to the decommissioning dynamics, the phase-out can be considered as beneficial for the economic viability of fossil power plants. Furthermore, the phase-out does not, overall, lead to environmental disadvantages in the medium term, but may be beneficial in the long run since lock-in effects are avoided.
Further research is required with regard to the consideration of future flexibility options and a new market design.Der steigende Anteil erneuerbarer Energien beeinflusst die Wirtschaftlichkeit von Investitionen in neue konventionelle Kraftwerke. Zahlreiche Studien haben diese Aspekte in Bezug auf den Kraftwerksbetrieb oder die langfristige Entwicklung des Kraftwerksparks untersucht. Stilllegungen, Investitionen und Betrieb im Kraftwerkspark bedingen jedoch einander. Aus diesem Grund wird in dieser Doktorarbeit ein Modellierungsansatz für eine integrierte Betrachtung von Kraftwerksstilllegung, -investition und -betrieb vorgestellt. In einer Fallstudie für Deutschland werden die Auswirkungen einer Integration erneuerbarer Energien auf Kraftwerksstilllegung, -investition und -betrieb im Zusammenhang mit unterschiedlichen Annahmen über die Restlaufzeit von Kernkraftwerken untersucht. Bezogen auf die Nutzung der Kernenergie wird hierbei ein Ausstiegsszenario sowie ein Laufzeitverlängerungsszenario (Verlän-gerung der Laufzeit um durchschnittlich 12 Jahre) betrachtet. Die Ergebnisse zeigen, dass die statische Stilllegung (d.h. die Betrachtung fester technischer Lebensdauern) im Fall eines Verzichts auf die Laufzeitverlängerung die im Kraftwerkspark verfügbare Leistung unterschätzt, da Retrofit-Maßnahmen (im Vergleich zur Stilllegung) nicht berücksichtigt werden. Die verfügbare Leistung im Falle einer Laufzeitverlängerung wird dagegen überschätzt, da die Möglichkeit der Kaltreserve (im Vergleich zum regulären Betrieb) vernachlässigt wird. Werden die Rückwirkungen der im Betrieb erwirtschaftbaren Deckungsbeiträge auf Stilllegungsentscheidungen (“dynamische Stilllegung”) betrachtet, so wird der strompreissenkende Effekt durch die Laufzeitverlängerung im Vergleich zur statischen Stilllegung mehr als halbiert. Knappheitssitutationen unterscheiden sich nicht wesentlich mit und ohne Laufzeitverlängerung im Fall der dynamischen Stilllegung, während bei statischer Stilllegung ohne Laufzeitzeitverlängerung ein deutlich größerer Importbedarf besteht. Die Fallstudie zeigt, dass weitere Systemflexibilitäten für die Integration erneuerbarer Energien benötigt werden. Der Anteil flexibler Kraftwerke ist größer im Fall des Kernenergieausstiegs. Der Kernenergieausstieg wirkt sich in Bezug auf die Stilllegungsdynamik positiv auf die Wirtschaftlichkeit fossiler Kraftwerke aus. Insgesamt führt der Kernenergieausstieg zu keinen mittelfristig nachteiligen Umwelteffekten, er kann sich jedoch langfristig positiv auswirken, da Lock-in-Effekte vermieden werden. Es besteht weiterer Forschungsbedarf in Bezug auf die Berücksichtigung künftiger Flexibilitätsoptionen und ein neues Marktdesign.