Sökning: "plutonium"
Visar resultat 21 - 25 av 38 avhandlingar innehållade ordet plutonium.
21. On the Diluent and Solvent Effects in Liquid-Liquid Extraction Systems based on Bis-triazine-bipyridine (BTBP) ligands
Sammanfattning : Used nuclear fuel is dangerous for mankind and her environment for a long time. If however, the minor actinides together with uranium and plutonium could be transmuted, i.e. transformed, into more shortlived or stable isotopes the volume of the waste could be significantly reduced together with a reduction in the radiotoxicity. LÄS MER
22. From the Electronic Structure of Point Defects to Functional Properties of Metals and Ceramics
Sammanfattning : Point defects are an inherent part of crystalline materials and they influence important physical and chemical properties, such as diffusion, hardness, catalytic activity and phase stability. Increased understanding of point defects enables us to tailor the defect-related properties to the application at hand. LÄS MER
23. Neutronic and burnup studies of accelerator-driven systems dedicated to nuclear waste transmutation
Sammanfattning : Partitioning and transmutation of plutonium, americium, and curium is inevitable if the radiotoxic inventory of spent nuclear fuel is to be reduced by more than a factor of 100. But, admixing minor actinides into the fuel severely degrades system safety parameters, particularly coolant void reactivity, Doppler effect, and (effective) delayed neutron fractions. LÄS MER
24. Why Faster is Better : On Minor Actinide Transmutation in Hard Neutron
Sammanfattning : In this thesis, options for efficient transmutation of transuranium elements are discussed. The focus is on plutonium, americium and curium mainly because of their long-term contribution to the radiotoxicity of spent nuclear fuel. Two innovative helium-cooled core designs are proposed, dedicated to the transmutation of actinides. LÄS MER
25. Generalised Campbell formulae for compound Poisson processes with applications in nuclear safeguards
Sammanfattning : Multiplicity counting is a widely used non-destructive assay method for estimating unknown parameters (primarily the mass) of samples of spontaneously fissioning materials (e.g. plutonium). Traditionally, measurements are performed with thermal neutron detectors operating in pulse counting mode. LÄS MER