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Made available in DSpace on 2014-10-09T12:47:41Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:08:15Z (GMT). No. of bitstreams: 1 08318.pdf: 5216839 bytes, checksum: a5aa91f30daca6826d66086676357dd7 (MD5) / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

The primary objective of this research is to develop an applied technology and provide an assessment for remotely measuring and analyzing the real time or near real time concentrations of used nuclear fuel (UNF) elements in electrorefiners (ER). Here, Laser-Induced Breakdown Spectroscopy (LIBS) in UNF pyroprocessing facilities was investigated. LIBS is an elemental analysis method, which is based on the emission from plasma generated by focusing a laser beam into the medium. This technology has been reported to be applicable in solids, liquids (includes molten metals), and gases for detecting elements of special nuclear materials. The advantages of applying the technology for pyroprocessing facilities are: (i) Rapid real-time elemental analysis; (ii) Direct detection of elements and impurities in the system with low limits of detection (LOD); and (iii) Little to no sample preparation is required. One important challenge to overcome is achieving reproducible spectral data over time while being able to accurately quantify fission products, rare earth elements, and actinides in the molten salt. Another important challenge is related to the accessibility of molten salt, which is heated in a heavily insulated, remotely operated furnace in a high radiation environment within an argon gas atmosphere. This dissertation aims to address these challenges and approaches in the following phases with their highlighted outcomes: 1. Aerosol-LIBS system design and aqueous testing: An aerosol-LIBS system was designed around a Collison nebulizer and tested using deionized water with Ce, Gd, and Nd concentrations from 100 ppm to 10,000 ppm. The average %RSD values between the sample repetitions were 4.4% and 3.8% for the Ce and Gd lines, respectively. The univariate calibration curve for Ce using the peak intensities of the Ce 418.660 nm line was recommended and had an R2 value, LOD, and RMSECV of 0.994, 189 ppm, and 390 ppm, respectively. The recommended Gd calibration curve was generated using the peak areas of the Gd 409.861 nm line and had an R2, LOD, and RMSECV of 0.992, 316 ppm, and 421 ppm, respectively. The partial least squares (PLS) calibration curves yielded similar results with RMSECV of 406 ppm and 417 ppm for the Ce and Gd curves, respectively. 2. High temperature aerosol-LIBS system design and CeCl3 testing: The aerosol-LIBS system was transitioned to a high temperature and used to measure Ce in molten LiCl-KCl salt within a glovebox environment. The concentration range studied was from 0.1 wt% to 5 wt% Ce. Normalization was necessary due to signal degradation over time; however, with the normalization the %RSD values averaged 5% for the mid and upper concentrations studied. The best univariate calibration curve was generated using the peak areas of the Ce 418.660 nm line. The LOD for this line was 148 ppm with the RMSECV of 647 ppm. The PLS calibration curve was made using 7 latent variables (LV) and resulting in the RMSECV of 622 ppm. The LOD value was below the expected rare earth concentration within the ER. 3. Aerosol-LIBS testing using UCl3: Samples containing UCl3 with concentrations ranging from 0.3 wt% to 5 wt% were measured. The spectral response in this range was linear. The best univariate calibration curves were generated using the peak areas of the U 367.01 nm line and had an R2 value of 0.9917. Here, the LOD was 647 ppm and the RMSECV was 2,290 ppm. The PLS model was substantially better with a RMSECV of 1,110 ppm. The LOD found here is below the expected U concentrations in the ER. The successful completion of this study has demonstrated the feasibility of using an aerosol-LIBS analytical technique to measure rare earth elements and actinides in the pyroprocessing salt.

京都大学 / 0048 / 新制・課程博士 / 博士(エネルギー科学) / 甲第22551号 / エネ博第402号 / 新制||エネ||77(附属図書館) / 京都大学大学院エネルギー科学研究科エネルギー基礎科学専攻 / (主査)教授 野平 俊之, 教授 萩原 理加, 教授 佐川 尚 / 学位規則第4条第1項該当 / Doctor of Energy Science / Kyoto University / DGAM

Le réacteur à sels fondus rapides (MSFR) est un des six concepts de réacteur nucléaire retenu lors du Forum Génération IV en 2001. La particularité de ce concept est d'utiliser un combustible liquide constitué d'un sel fondu, LiF-ThF₄-UF₄/UF3₃ (77-19-4 mol%) et d'intégrer un procédé de traitement du sel usé. Ce traitement est constitué d'étapes successives de séparation chimiques basées sur les propriétés redox et acido-basiques des éléments produits dans le réacteur par des réactions nucléaires : produits de fission solubles et gazeux, éléments métalliques et actinides mineurs solubles. L'une des étapes majeures du procédé de traitement est une extraction réductrice qui consiste à mettre en contact le sel fondu et un métal liquide, le bismuth, contenant un élément réducteur, le lithium. Cette étape permet notamment de séparer les actinides mineurs des lanthanides. Les actinides mineurs sont réintroduits dans le réacteur nucléaire afin d'y être brûler alors que les lanthanides seront confinés en stockage profond.Le travail réalisé au cours de cette thèse avait deux objectifs : (i) vérifier la faisabilité de l'extraction réductrice des actinides et des lanthanides, étape qui avait été validée au préalable uniquement sur la base de calculs thermodynamiques et (ii) étudier la chimie des sels fluorures fondus (et notamment le sel combustible LiF-ThF₄-UF₄) en développant une méthodologie pour la détermination de données fondamentales telles que les coefficients d'activité dans les milieux fluorures, coefficients qui quantifient les propriétés de solvatation.La première étape pour réaliser expérimentalement une extraction réductrice consiste à préparer une nappe métallique de Bi-Li liquide de composition pré-définie. Une technique d'électrolyse en milieu LiCl-LiF fondu à 550°C a été retenue pour réaliser ces solutions métalliques. Nous avons montré que seul ce milieu fondu pouvait être utilisé pour la fabrication de ces alliages métalliques. Des tests d'extraction ont ensuite été réalisés par contact entre LiF-ThF₄ (dans lequel sont introduits UF₄ et NdF ₃ pour simuler respectivement les actinides et les lanthanides) et Bi-Li à 650°C. Les principaux résultats montrent que l'extraction du néodyme et de l'uranium a été obtenue avec des rendements respectivement de l'ordre de 3% et 15% dans les meilleures conditions. Ces valeurs sont faibles comparées aux calculs thermodynamiques prévisionnels. On explique la faible efficacité de l'extraction par une extraction simultanée du thorium dans la nappe métallique liquide qui forme des composés intermétalliques à l'interface métal/sel et bloque le transfert interphasique. Des méthodes ont été développées pour atteindre des données fondamentales qui font défaut en milieu fluorures fondus, en particulier les propriétés de solvatation. La spéciation de plusieurs cations métalliques par les ions fluorures à haute température a notamment été étudiée et les constantes de complexation calculées par simulation des résultats expérimentaux. Réalisée pour deux lanthanides, le néodyme et le lanthane, deux actinides, le thorium et l'uranium et également pour un métal de transition, le nickel, cette étude permet d'atteindre les coefficients d'activité de ces éléments dans tous les sels fluorures fondus. En particulier, l'étude de la spéciation du thorium a été une étape importante dans la connaissance de la chimie du sel combustible LiF-ThF₄ puisque nous avons pu en déduire le coefficient d'activité de l'ion fluorure dans ce milieu à 650°C.Enfin, l'ensemble de ce travail a conduit à donner une première estimation de la réactivité de chaque élément de la classification périodique (présent dans le réacteur nucléaire après opération) à chaque étape du traitement du sel combustible usé. / The molten salt fast reactor (MSFR) is one of the six nuclear reactor concepts retained during the Forum GEN IV in 2001. The particularity of this concept is to use a liquid fuel consisting of a molten salt, LiF-ThF₄-UF₄ /UF ₃ (77-19-4 mol%) and to have an integrated spent fuel treatment process. This treatment consists of successive chemical separation steps based on redox and acid-base properties of the elements produced in the reactor by nuclear reactions: soluble and gaseous fission products, metals elements and soluble minor actinides. One of the major steps of the treatment method is a reducing extraction which consists to contact the molten salt and a liquid metal, bismuth, containing the reducing element, lithium. This step allows separating the minor actinides and lanthanides. Minor actinides are reintroduced in the nuclear reactor to be burned while the lanthanides are confined in deep storage.The work in this thesis had two objectives: (i) assess the feasibility of reducing extraction of actinides and lanthanides, a step that had previously only been validated on the basis of thermodynamic calculations and (ii) study the chemistry of molten fluoride salts (and especially the fuel salt) by developing a methodology for the determination of fundamental data such as the activity coefficients in fluorides media, coefficients activities which quantify the solvation properties.To experimentally realize a reducing extraction, the first step is to prepare a metal layer of liquid Bi-Li with predefined composition. An electrolysis technique in molten salt LiCl-LiF at 550°C was chosen to achieve these metal solutions. We have shown that only this molten medium could be used for the manufacture of such metal alloys. Extraction tests were then carried out by contact between LiF-ThF₄ (with UF₄ and NdF ₃ are introduced to simulate respectively the actinides and lanthanides) and Bi-Li at 650°C. The main results show that the extraction of neodymium and uranium was obtained with yields of around 3% and 15% respectively in the best conditions. These values are low compared to previous thermodynamic calculations. Low efficiency of the extraction is due to a simultaneous extraction of thorium in the liquid metal phase which forms intermetallic compounds at the metal/salt interphase and blocks the transfer.Methods have been developed to achieve fundamental data that are lacking in molten fluoride medium, in particularly the solvation properties. Speciation of some metallic cations by fluoride ions with high temperature was particularly studied and calculation of complexation constants by simulated experimental results was done. Carried out for two lanthanides, neodymium and lanthanum, two actinides, thorium and uranium, and also for a transition metal, nickel, this study achieves to calculate the activity coefficients of these elements in different fluoride molten salt. The study of the speciation of thorium was an important step to understand the chemistry of the fuel salt LiF-ThF₄. We were able to calculate the activity coefficient of the fluoride ion in this environment at 650°C.Finally, all of this work allows giving a first estimate of the reactivity of each element of the periodic table (present in the nuclear reactor after operation) at each stage of the treatment of the spent fuel salt.

The present work aims to develop Aluminium metal matrix composites by incorporating of reinforcements such that combination of best properties could be achieved. The metal base was selected was Aluminium 6082 and it is reinforced with varying volume percentage of Alumina oxide and Fly ash. These AMC were developed by using stir casting technique, in which predetermined reinforcement is added to the molten matrix is stirred well to obtain desired castings. These castings were studied for behaviour and subjected to mechanical testing to study the effects of various reinforcements. The Rockwell hardness and tensile tests revealed that composite with 7.5 % Al2O3 5% Fly ash shows highest hardness value of 69 HRB and 100.370 N/mm2, which is better than base alloy as well as impact is more without reinforcement ratio shows highest impact strength value 9 joules. Highly reinforced composites show higher variations due to the agglomeration of particles.

No description available.

No description available.

Nuclear fusion is considered a clean energy source: it emits no CO2 and leaves little radioactive waste. It is important to start paving the path toward nuclear fusion whilst simultaneously moving away from fossil fuels and carbon emissions. One of the challenges of nuclear fusion is the lack of tritium, which, together with deuterium makes up its fuel. This research is focused on utilizing one current method of nuclear fission technology, namely molten salt reactors, to generate at least the initial loads of tritium for the first fusion reactors. Current research is primarily focused on providing tritium during the nuclear fusion reaction. However, it is also necessary to have a tritium supply whenever we start up a nuclear fusion reactor. The largest source of tritium is the CANDU nuclear fission reactor. A typical 500 MW CANDU produces 130 g of tritium annually as a biproduct of power generation. However, a future commercial fusion power plant is expected to consume 300 g of tritium per day to produce 800 MW. Thus, this research explores the possibility of breeding tritium in other fission reactors, in particular molten salt reactors (MSR). MCNP4C was used to simulate a simple Molten Salt Reactor setting with 61 molten salt fuel channels and applying a molten salt blanket to study how the presence of specific elements in the blanket affects tritium production, as well as criticality. The study relies on nuclear data from the National Nuclear Data Center (NNDC), and Oak Ridge National Laboratory (ORNL) as benchmark to verify the accuracy of the results. The calculated output of tritium is 325 g/year for a 100 MW (th) reactor, which is considered a positive outcome that opens the door for more research in this direction. / Thesis / Master of Applied Science (MASc)

The next generation of nuclear power sources, Gen. IV, will include an emphasis on small, modular reactor (SMR) designs, which will allow for standardized, factory-based manufacturing and flexibility in the design of power plants by utilizing one or several modular reactor units in parallel. One of the reactor concepts being investigated is the Molten Salt Reactor concept (MSR), which utilizes a molten salt flow loop to cool the reactor and transfer heat to the power conversion cycle (PCS).Here, the use of a supercritical carbon dioxide (S-CO2) Brayton cycle is assumed for that PCS. The purpose of this thesis is to investigate the heat exchange between these two systems and to determine the suitability of a common heat exchanger concept, the shell-and-tube heat exchanger (STHE). This was accomplished using a code written in Python programming language that optimized the geometry ofa baffled STHE for a range of conditions the reflect MSR power plants currently in the design or concept stages. Star-CCM+ computational fluid dynamics (CFD)software was used to visualize the flow patterns of molten salt and CO2 in these STHE designs, and it was also used to determine heat transfer coefficients and pressure drops. These values were compared to those calculated by the optimizer code in order to validate its results. Finally, modularity analysis was performed for these STHE designs. Trends were generalized from these results that will contribute to judgments about the suitability of STHE’s for use with MSR’s and S-CO2.