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  • About
  • The Global ETD Search service is a free service for researchers to find electronic theses and dissertations. This service is provided by the Networked Digital Library of Theses and Dissertations.
    Our metadata is collected from universities around the world. If you manage a university/consortium/country archive and want to be added, details can be found on the NDLTD website.
21

Etude de la cinétique et du mécanisme de la réaction de dissolution du bioxyde de plutonium par l'ion Cr (II) en solution acide /

Machuron-Mandard, Xavier. January 1991 (has links)
Th. Univ.--Chimie anal.--Paris 6, 1991. / Résumé en anglais. Bibliogr. p. 174-180.
22

Deuteron-induced spallation and fission reactions in plutonium isotopes

Luoma, Ernie Victor. January 1956 (has links)
Thesis (M.S. in Chemistry)--University of California, Berkeley, Novmber 1956. / Includes bibliographical references (leaves 34-35).
23

Influence of temperature on the extraction of Pu(IV) by tri-n-butyl phosphate from acidic nitrate solutions /

Brown, M. Alex. January 1900 (has links)
Thesis (M.S.)--Oregon State University, 2010. / Printout. Includes bibliographical references (leaves 43-45). Also available on the World Wide Web.
24

Plutonium isotopes in the North Atlantic.

Buesseler, Ken O. January 1986 (has links)
Thesis (Ph. D.)--Massachusetts Institute of Technology and the Woods Hole Oceanographic Institution, 1986. / Includes bibliographical references (p. 193-207).
25

Microstructural and morphological aspects of plutonium hydride

Brierley, Martin January 2016 (has links)
Plutonium is a hazardous radioactive material; the α-particles that are emitted are particularly damaging to health should contamination be inhaled or ingested into the body. During long term storage a number of conditions have been observed which can cause plutonium to corrode, which liberates particles from the surface. It is imperative to understand the processes involved in the corrosion of plutonium during long term storage to predict the likely state that metallic pieces may be found should subsequent handling be required. The growth mechanisms of plutonium hydride beyond the nucleation stage are not well understood. Detailed characterisation of the microstructural features associated with hydride reaction sites is required to develop a mechanistic understanding of the growth stage of hydrogen corrosion. Suitable processes and analysis methods were developed using cerium as an analogous material to δ-plutonium; during this stage, the knowledge of the corrosion of cerium by hydrogen was significantly improved using in situ gas dosing equipment, metallographic preparation, light microscopy, scanning electron microscopy (SEM), ion milling, secondary ion mass spectrometry (SIMS), atomic force microscopy (AFM) and vacuum nanoindentation. SEM and ion milling methods developed on cerium were subsequently used on the analysis on pre-formed and passivated hydride reaction sites on δ-plutonium .In situ exposure of electro-refined plutonium and a Pu 0.3 wt% Ga alloy were investigated without prior exposure to oxygen, revealing the as-formed microstructure of the hydride reaction product to be analysed. Subsequent metallographic preparation was used to confirm findings from the in situ analysis. The highest resolution analysis of the hydride product formed on cerium, delta plutonium and electro-refined plutonium has been obtained to date. Hydride reaction sites formed on cerium and δ-Pu were observed to be oblate, confirming growth anisotropy. A mechanism for the anisotropic growth was proposed where the stress fields introduced into the metal surrounding a lower density hydride play a significant role in further development of a hydride reaction site, causing failure of the surface oxide diffusion barrier surrounding a hydride reaction site.
26

Dissolution réductrice d'oxydes de lanthanides et de PuO2 assistée par ultrasons / Reductive dissolution of lanthanide oxides and PuO2 assisted by ultrasound

Beaudoux, Xavier 23 January 2015 (has links)
Dans le cadre du programme nucléaire français, le combustible des réacteurs est constitué d'oxydes d'uranium ou d'oxydes mixtes d'uranium et de plutonium (appelé MOX). Des développements sont constamment effectués sur les procédés de retraitement de ces combustibles afin d'optimiser la récupération des matières valorisables et de minimiser le volume des déchets. Lors de la dissolution des MOX, la quantité de fines de dissolution résiduelles riches en Pu est parfois importante malgré l'utilisation de conditions chimiques dures (oxydantes et corrosives). La difficulté à dissoudre des lots de PuO2 déclarés non conformes lors de la fabrication du MOX peut également représenter un verrou technologique. Dans ce contexte, la sonochimie est envisagée comme alternative aux méthodes actuelles de dissolution de PuO2, ou de MOX enrichi en Pu. Dans un premier temps, des travaux de dissolution sonochimique ont été réalisés sur un analogue inactif de PuO2, à savoir CeO2. Les résultats obtenus ont ainsi permis d'orienter les expériences de dissolution de PuO2. En conditions réductrices et acides, beaucoup plus douces que celles utilisées industriellement, la dissolution totale de ces deux oxydes a été effectuée en quelques heures. Parallèlement, une étude connexe a montré qu'il est possible de dissoudre totalement des oxydes mixtes de lanthanides à base de Ce, par un procédé de dissolution sonocatalytique et réductrice en présence de nanoparticules de Pt. La dissolution est d'autant plus rapide que la teneur en lanthanides trivalents est importante au sein de l'oxyde. Enfin, une dernière partie a été consacrée à la dissolution sous agitation magnétique d'oxydes à base de Ce en présence ou non de métaux nobles, dans des milieux faiblement acides contenant des molécules naturelles réductrices. Dans ces conditions, une dissolution totale, rapide et sélective de ces oxydes a été observée. Ces deux dernières études présentent un intérêt dépassant le cadre du nucléaire et applicable au recyclage de matériaux industriels (pots catalytiques, piles à combustible…). / In the French nuclear program, the reactor fuel consists of uranium oxides or uranium plutonium mixed oxides (called MOX). Developments are constantly made on the resulting reprocessing of these fuels in order to optimize the recovery of reusable materials and to minimize the waste volume. In the case of MOX dissolution, the amount of Pu-rich dissolution residues is sometimes high despite the use of hard chemical conditions (oxidizing and corrosive). The difficulty to dissolve PuO2 batches declared non-standard during the fabrication of MOX can also be a technological barrier. In this context, sonochemistry can be considered as an alternative to current methods of dissolution of PuO2 or Pu enriched MOX. First, experiments of sonochemical dissolution were performed on an inactive analogue of PuO2, namely CeO2. The results were then used as a working basis for the dissolution of PuO2. Under reducing and acidic conditions, much milder than those used industrially, the complete dissolution of these two oxides was carried out within a few hours. Meanwhile, a related study showed that it is possible to completely dissolve lanthanide mixed oxides by a process of sonocatalytic and reductive dissolution in the presence of Pt. The dissolution rates increase with the trivalent lanthanide content within the oxide. Finally, the last part was devoted to the dissolution under magnetic stirring of Ce-based oxides in the presence or absence of noble metals, in weakly acidic media containing reducing natural molecules. Under these conditions, a complete, rapid and selective dissolution of these oxides was observed. These last two studies present an interest beyond the scope of nuclear chemistry, concerning the recycling of industrial materials (catalytic converters, fuel cells...).
27

An investigation of the performance characteristics of isothermal calorimeters

Jones, Andrew Christopher January 1997 (has links)
No description available.
28

Development of a portable neutron coincidence counter for field measurements of nuclear materials using the advanced multiplicity capabilities of MCNPX 2.5.F and the neutron coincidence point model

Thornton, Angela Lynn 10 October 2008 (has links)
Neutron coincidence counting is an important passive Nondestructive Assay (NDA) technique widely used for qualitative and quantitative analysis of nuclear material in bulk samples. During the fission process, multiple neutrons are simultaneously emitted from the splitting nucleus. These neutron groups are often referred to as coincident neutrons. Because different isotopes possess different coincident neutron characteristics, the coincident neutron signature can be used to identify and quantify a given material. In an effort to identify unknown nuclear samples in field inspections, the Portable Neutron Coincidence Counter (PNCC) has been developed. This detector makes use of the coincident neutrons being emitted from a bulk sample. An in-depth analysis has been performed to establish whether the nuclear material in an unknown sample could be quantified with the accuracy and precision needed for safeguards measurements. The analysis was performed by comparing experimental measurements of PuO2 samples to the calculated output produced using MCNPX and the Neutron Coincidence Point Model. Based on the analysis, it is evident that this new portable system can play a useful role in identifying nuclear material for verification purposes.
29

Development of a portable neutron coincidence counter for field measurements of nuclear materials using the advanced multiplicity capabilities of MCNPX 2.5.F and the neutron coincidence point model

Thornton, Angela Lynn 15 May 2009 (has links)
Neutron coincidence counting is an important passive Nondestructive Assay (NDA) technique widely used for qualitative and quantitative analysis of nuclear material in bulk samples. During the fission process, multiple neutrons are simultaneously emitted from the splitting nucleus. These neutron groups are often referred to as coincident neutrons. Because different isotopes possess different coincident neutron characteristics, the coincident neutron signature can be used to identify and quantify a given material. In an effort to identify unknown nuclear samples in field inspections, the Portable Neutron Coincidence Counter (PNCC) has been developed. This detector makes use of the coincident neutrons being emitted from a bulk sample. An in-depth analysis has been performed to establish whether the nuclear material in an unknown sample could be quantified with the accuracy and precision needed for safeguards measurements. The analysis was performed by comparing experimental measurements of PuO2 samples to the calculated output produced using MCNPX and the Neutron Coincidence Point Model. Based on the analysis, it is evident that this new portable system can play a useful role in identifying nuclear material for verification purposes.
30

Development of a portable neutron coincidence counter for field measurements of nuclear materials using the advanced multiplicity capabilities of MCNPX 2.5.F and the neutron coincidence point model

Thornton, Angela Lynn 15 May 2009 (has links)
Neutron coincidence counting is an important passive Nondestructive Assay (NDA) technique widely used for qualitative and quantitative analysis of nuclear material in bulk samples. During the fission process, multiple neutrons are simultaneously emitted from the splitting nucleus. These neutron groups are often referred to as coincident neutrons. Because different isotopes possess different coincident neutron characteristics, the coincident neutron signature can be used to identify and quantify a given material. In an effort to identify unknown nuclear samples in field inspections, the Portable Neutron Coincidence Counter (PNCC) has been developed. This detector makes use of the coincident neutrons being emitted from a bulk sample. An in-depth analysis has been performed to establish whether the nuclear material in an unknown sample could be quantified with the accuracy and precision needed for safeguards measurements. The analysis was performed by comparing experimental measurements of PuO2 samples to the calculated output produced using MCNPX and the Neutron Coincidence Point Model. Based on the analysis, it is evident that this new portable system can play a useful role in identifying nuclear material for verification purposes.

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