Si le cycle du thorium possède plusieurs avantages par rapport au cycle U/Pu, notamment une meilleure régénération de la matière fissile en spectre thermique et une production moindre d'actinides mineurs, il présente plusieurs limites. L'une d'elles est la présence, dans le combustible thorié irradié, d'232U. Cet isotope est le précurseur d'un rayonnement γ de 2.6 MeV. Cette thèse a, en partie, pour objectif d'étudier les différents paramètres influençant la synthèse de ce noyau dans divers types de combustibles et de réacteurs.L'autre partie de ce travail consiste à estimer l'impact de cet indésirable sur la radioprotection de l'aval du cycle. Dans ce but, un ensemble d'outils, permettant le calcul des spectres énergétiques des différents rayonnements émis par la matière radioactive, a été spécialement développé. Ces outils, dont la véracité a été éprouvée par l'intermédiaire de plusieurs benchmarks, fait partie intégrante de ce travail de thèse. / The thorium cycle is a good candidate for replacing the current U/Pu cycle since the fissile nucleus of the cycle, 233U, has neutronic properties favourable to a much better regeneration of fissile material in thermal reactors. Moreover, the production of minor actinides is significantly reduced. However, the use of Thorium is only viable if the spent fuel is reprocessed to recover the fissile 233U that does not exist in nature. This reprocessing will involve a heavy industrial infrastructure, particularly since thorium based spent fuel contains small quantities of 232U that is the mother of the hard gamma emitter (208Tl) of 2.6 MeV. The goal of this thesis is, firstly, to study the parameters related to the synthesis of 232U in several kind of fuels and reactors. In a second part, the thesis focuses on the impact on radioprotection of the back end of the fuel in case of switching from the current uranium (U/Pu) cycle to the thorium (Th/U) cycle. For this last purpose, CHARS (CHAracterization of Radioactives Sources) was developed during this thesis. This code, validated by several benchmarks, handles the calculation of radiation sources in all aspects of the fuel cycle.

Master of Science

The Daya Bay experiment has determined the last unknown mixing angle $theta_{13}$. This thesis describes the layout of the experiment and the detector design. The analysis presented in the thesis covered the water attenuation, spent fuel neutrino and electron anti-neutrino spectrum. Other physics analysis and impact to future experiments are also discussed. / Ph. D.

Pre-detonation technical nuclear forensics techniques for research reactor spent fuel were developed in a collaborative project with Savannah River National Lab ratory. An inverse analysis method was employed to reconstruct reactor parameters from a spent fuel sample using results from a radiochemical analysis. In the inverse analysis, a reactor physics code is used as a forward model. Verification and validation of different reactor physics codes was performed for usage in the inverse analysis. The verification and validation process consisted of two parts. The first is a variance analysis of Monte Carlo reactor physics burnup simulation results. The codes used in this work are MONTEBURNS and MCNPX/CINDER. Both utilize Monte Carlo transport calculations for reaction rate and flux results. Neither code has a variance analysis that will propagate through depletion steps, so a method to quantify and understand the variance propagation through these depletion calculations was developed. The second verification and validation process consisted of comparing reactor physics code output isotopic compositions to radiochemical analysis results. A sample from an Oak Ridge Research Reactor spent fuel assembly was acquired through a drilling process. This sample was then dissolved in nitric acid and diluted in three different quantities, creating three separate samples. A radiochemical analysis was completed and the results were compared to simulation outputs at different levels ofdetail. After establishing a forward model, an inverse analysis was developed to re-construct the burnup, initial uranium isotopic compositions, and cooling time of a research reactor spent fuel sample. A convergence acceleration technique was used that consisted of an analytical calculation to predict burnup, initial 235U, and 236U enrichments. The analytic calculation results may also be used stand alone or in a database search algorithm. In this work, a reactor physics code is used as a for- ward model with the analytic results as initial conditions in a numerical optimization algorithm. In the numerical analysis, the burnup and initial uranium isotopic com- positions are reconstructed until the iterative spent fuel characteristics converge with the measured data. Upon convergence of the sample’s burnup and initial uranium isotopic composition, the cooling time can be reconstructed. To reconstruct cooling time, the standard decay equation is inverted and solved for time. Two methods were developed. One method uses the converged burnup and initial uranium isotopic compositions along in a reactor depletion simulation. The second method uses an isotopic signature that does not decay out of its mass bin and has a simple production chain. An example would be 137Cs which decays into the stable 137Ba. Similar results are achieved with both methods, but extended shutdown time or time away from power results in over prediction of the cooling time. The over prediction of cooling time and comparison of different burnup reconstruction isotope results are indicator signatures of extended shutdown or time away from power. Due to dynamic operation in time and function, detailed power history reconstruction for research reactors is very challenging. Frequent variations in power, repeated variable shutdown time length, and experimentation history affect the spectrum an individual assembly is burned with such that full reactor parameter reconstruction is difficult. The results from this technical nuclear forensic analysis may be used with law enforcement, intelligence data, macroscopic and microscopic sample characteristics in a process called attribution to suggest or exclude possible sources of origin for a sample.

Three models have been developed and applied in the performance assessment of a final repository. They are based on accepted theories and experimental results for known and possible mechanisms that may dominate in the oxidative dissolution of spent fuel and the release of nuclides from a canister. Assuming that the canister is breached at an early stage after disposal, the three models describe three sub-systems in the near field of the repository, in which the governing processes and mechanisms are quite different. In the model for the oxidative dissolution of the fuel matrix, a set of kinetic descriptions is provided that describes the oxidative dissolution of the fuel matrix and the release of the embedded nuclides. In particular, the effect of autocatalytic reduction of hexavalent uranium by dissolved H2, using UO2 (s) on the fuel pellets as a catalyst, is taken into account. The simulation results suggest that most of the radiolytic oxidants will be consumed by the oxidation of the fuel matrix, and that much less will be depleted by dissolved ferrous iron. Most of the radiolytically produced hexavalent uranium will be reduced by the autocatalytic reaction with H2 on the fuel surface. It will reprecipitate as UO2 (s) on the fuel surface, and thus very little net oxidation of the fuel will take place. In the reactive transport model, the interactions of multiple processes within a defective canister are described, in which numerous redox reactions take place as multiple species diffuse. The effect of corrosion of the cast iron insert of the canister and the reduction of dissolved hexavalent uranium by ferrous iron sorbed onto iron corrosion products and by dissolved H2 are particularly included. Scoping calculations suggest that corrosion of the iron insert will occur primarily under anaerobic conditions. The escaping oxidants from the fuel rods will migrate toward the iron insert. Much of these oxidants will, however, be consumed by ferrous iron that comes from the corrosion of iron. The nonscavenged hexavalent uranium will be reduced by ferrous iron sorbed onto the iron corrosion products and by dissolved hydrogen. In the transport resistance network model, the transport of reactive actinides in the near field is simulated. The model describes the transport resistance in terms of coupled resistors by a coarse compartmentalisation of the repository, based on the concept that various ligands first come into the canister and then diffuse out to the surroundings in the form of nuclide complexes. The simulation results suggest that carbonate accelerates the oxidative dissolution of the fuel matrix by stabilizing uranyl ions, and that phosphate and silicate tend to limit the dissolution by the formation of insoluble secondary phases. The three models provide powerful tools to evaluate "what if" situations and alternative scenarios involving various interpretations of the repository system. They can be used to predict the rate of release of actinides from the fuel, to test alternative hypotheses and to study the response of the system to various parameters and conditions imposed upon it. / QC 20100521

Submitted by Marco Antonio Oliveira da Silva (maosilva@ipen.br) on 2016-08-26T12:33:02Z No. of bitstreams: 0 / Made available in DSpace on 2016-08-26T12:33:02Z (GMT). No. of bitstreams: 0 / Este trabalho tem como objetivo estudar as modificações introduzidas, ao longo de sucessivas versões, nos modelos empíricos do programa computacional FRAPCON utilizado para a simulação do comportamento sob irradiação de varetas combustíveis de Reatores a Água Leve Pressurizada (Pressurized Water Reactor - PWR) em regime de estado estacionário e sob condições de alta queima. No estudo, foram analisados os modelos empíricos utilizados pelo FRAPCON e que são apresentados em sua documentação oficial. Um estudo bibliográfico foi conduzido sobre os efeitos da alta queima em combustíveis nucleares visando melhorar o entendimento dos modelos utilizados pelo FRAPCON nestas condições. Foram feitas simulações do comportamento sob irradiação de uma vareta combustível típica de um reator PWR utilizando as versões 3.3, 3.4 e 3.5 do FRAPCON. Os resultados apresentados pelas diferentes versões do programa foram comparados entre si de forma a verificar as consequências das mudanças de modelo nos parâmetros de saída do programa. Foi possível observar que as modificações introduzidas trouxeram diferenças significativas nos resultados de parâmetros térmicos e mecânicos da vareta combustível, principalmente quando se evoluiu da versão FRAPCON-3.3 para a versão FRAPCON-3.5. Nessa ultima versão, obteve-se menores temperaturas na vareta combustível, menores tensões e deformações no revestimento, menor espessura da camada de oxido formada no revestimento a altas queimas na vareta combustível. / Dissertação (Mestrado em Tecnologia Nuclear) / IPEN/D / Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP

A geological repository for spent nuclear fuel (SNF) and an associated encapsulation facility will be built in Sweden.  The encapsulation facility is planned to be in operation in 2025 and it will be the last place where verifying safeguards measurements of SNF can be performed. It is not clear what types of measurements that will be performed, because such requirements are not yet posed by national and international authorities and inspecting organizations. This report describes the objective and most recent results of a master thesis project, whereby a few existing non-destructive assay techniques for verifying SNF are selected for a review. The study focuses on the verifying ability of different techniques, or system of techniques in relation to the requirement that may be put on the future encapsulation plant. In addition, possible needs for future simulations and measurements are discussed. The work is done as a collaboration between Uppsala University in Sweden and Los Alamos National Laboratory in the USA.

The PTR system allows the EPR2 fuel pool to be cooled. The evacuation of the residual power fromthe pool is ensured by several heat exchangers and pumps, which have to be dimensioned in order to meetdifferent requirements.In order to dimension them, the worst-case scenario of the components must first be determined.Sensitivity to external conditions and efficiency studies enable to propose a heat exchanger design tomeet the requirements. A parametric study then allows to study more precisely the influence of thegeometry of the exchanger on the heat transfer. This allows to guide the conception of a CFD study ofthe design on the Comsol software in order to validate it. The proposed design can then be integratedinto the PTR cooling train. The train is modeled with FloMaster, in order to compute the head losses inthe hydraulic system and to propose a pump altimetry preventing cavitation.The dimensioning case of the exchangers corresponds to the operating case of the PTR trains duringunit shutdown, while the scenario that facilitates cavitation corresponds to the boiling of the fuel pool.The temperature of the cold source RRI is a sensitive data for the operation of the exchangers. In addition,the placement of the baffles and the space between the tubes play a determining role in the heat removal.It was difficult to construct the desired exchanger geometry in CFD. A compromise model was thusidentified and studied in CFD. The FloMaster study showed that the pressure drop in the PTR network isabout 15.5 mCE at the considered flow rate. Cavitation in a main train is not a problem if the pumps arelowered by at least 1.8 meters from the pool suction point.The sizing study therefore allowed us to propose a heat exchanger design close to the specifications,but this could not be precisely studied in CFD. The pressure drop study allowed to propose a pumpaltimetry preventing cavitation. / PTR-systemet gör det möjligt att kyla bränslebassängen i EPR2. Evakueringen av den återstående energin frånfrån bassängen säkerställs av flera värmeväxlare och pumpar, som måste dimensioneras för att uppfyllaolika krav.För att kunna dimensionera dem måste man först fastställa det värsta tänkbara scenariot för komponenterna.Känslighet för yttre förhållanden och effektivitetsstudier gör det möjligt att föreslå en värmeväxlardesign somuppfyller kraven. En parametrisk studie gör det sedan möjligt att mer exakt studera påverkan avväxlarens geometri har på värmeöverföringen. Detta gör det möjligt att vägleda utformningen av en CFD-studie avav konstruktionen i programvaran Comsol för att validera den. Den föreslagna konstruktionen kan sedan integrerasi PTR-kyltåget. Tåget modelleras med FloMaster, för att beräkna huvudförlusterna ihydraulsystemet och för att föreslå en pumphöjdmätning som förhindrar kavitation.Dimensioneringsfallet för växlarna motsvarar driftsfallet för PTR-tågen under driftavställning avenhetens avstängning, medan det scenario som underlättar kavitation motsvarar kokning av bränslebassängen.Temperaturen hos den kalla källan RRI är en känslig uppgift för driften av växlarna. Dessutom måsteplaceringen av bafflarna och utrymmet mellan rören en avgörande roll för värmeavledningen.Det var svårt att konstruera den önskade växlargeometrin i CFD. En kompromissmodell identifierades därföridentifierades och studerades i CFD. FloMaster-studien visade att tryckfallet i PTR-nätverket ärcirka 15,5 mCE vid det aktuella flödet. Kavitation i ett huvudtåg är inte ett problem om pumparna ärsänks med minst 1,8 meter från poolens sugpunkt.Dimensioneringsstudien gjorde det därför möjligt för oss att föreslå en värmeväxlardesign som ligger nära specifikationerna,men detta kunde inte studeras exakt i CFD. Tryckfallsstudien gjorde det möjligt att föreslå en pumpaltimetri som förhindrar kavitation.

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Made available in DSpace on 2014-10-09T12:43:03Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:05:39Z (GMT). No. of bitstreams: 1 06173.pdf: 6069998 bytes, checksum: 60ab3760f99f6d97fd52766b4d449ab5 (MD5) / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP