Made available in DSpace on 2014-10-09T12:49:28Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:01:16Z (GMT). No. of bitstreams: 0 / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares, IPEN/CNEN-SP

Made available in DSpace on 2014-10-09T12:49:28Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:01:16Z (GMT). No. of bitstreams: 0 / Quando um país se torna auto-suficiente em uma parte do ciclo nuclear, quanto à produção de combustível que será usado em suas centrais nucleares para a geração de energia, precisa voltar sua atenção para a melhor forma de armazenar este combustível após a sua utilização. A armazenagem do combustível nuclear queimado é uma prática necessária e utilizada nos dias atuais em todo o mundo como temporária, tanto por países que não têm definido o plano de destinação final, isto é, o repositório definitivo, como também por aqueles que já o possuem. Existem dois aspectos principais que envolvem os combustíveis queimados: um referente à armazenagem do combustível nuclear queimado destinado ao reprocessamento e o outro ao que será enviado para deposição final quando o sítio de deposição definitiva estiver definido, corretamente localizado, adequadamente caracterizado quanto aos diversos aspectos técnicos, e licenciado. Este último aspecto pode envolver décadas de estudos por causa das definições técnicas e normativas em um dado país. No Brasil, o interesse está voltado para a armazenagem dos combustíveis queimados que não serão reprocessados. Este trabalho analisa os tipos possíveis de armazenagem, o panorama internacional e a possível proposta para a futura construção de um sítio de armazenagem temporária no país. / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares, IPEN/CNEN-SP

Thesis (M.S.)--University of Nevada, Reno, 2007. / "May, 2007." Includes bibliographical references. Online version available on the World Wide Web.

The continual increase in electricity dependence for the advancement of society has led to increased demand in electricity globally. This increased demand, among other things such as global warming interventions and energy security have encouraged the need to diversify electricity generation sources. Civilian use of nuclear power dates back to the 1950s. The United States of America and France are currently leading with the highest nuclear power generation in the world, generating 101 GWe and 63 GWe, respectively. Several countries such as China and the United Arab Emirates have committed to new nuclear build in order to increase their nuclear power generation capacities. Standing against the prospects of growth of the nuclear power industry are technical and nontechnical challenges. These include proliferation risk, safety, high capital costs and high-level waste management. Most spent nuclear fuel from power reactors is currently stored in the spent fuel pools on reactor sites, and some have been reprocessed. It is estimated that about 32% (370 000 tons of Heavy Metal) of the total spent fuel generated from power reactors have been reprocessed up to date. With most of the spent fuel pools filling up, alternative interim and long term disposal of spent nuclear fuel solutions have been under investigation from as early as the 1970s. South Africa has planned an interim dry storage facility for the spent nuclear fuel to be established at the existing Koeberg power station. The interim dry storage facility will make use of HI-STAR 100 multi-purpose casks to store spent nuclear fuel until the country decides on final disposal solution. There are many aspects that are critical to safe, efficient and cost-effective long term storage of spent nuclear fuel. Some of the physics and engineering aspects concerning dry storage facilities are briefly discussed. The aspects presented here are: radiation containment, spent fuel, sub-criticality, decay heat removal, site location aspects, response to seismic events, cask corrosion, transportation infrastructure, operability and monitoring. The study of the three existing dry cask storages from the USA, Hungary and Belgium gives an overview of the dry cask technology in use today. These presentations are based on publicly available reliable information. The proposed dry storage facility at Koeberg will be in the existing power station footprint using the HI-STAR 100 casks. The decision to have the proposed dry storage facility at Koeberg will minimise related licence applications and part of security installations as the site already has some security. The location of the facility in the power station’s footprint also allows for cost-effective and safe transportation of casks from the reactor building to the proposed facility. The modularity aspect of the dry cask storage facility at MV Paks in Hungary should also be employed at Koeberg to allow for more storage. This will cater for additional casks that may need to be stored if more nuclear power plants are procured in the future. South Africa’s air traffic around the Western Cape is not as congested as Belgium’s. There is, therefore, no need for the casks to be housed in concrete buildings like Doel’s. Most of Koeberg’s high-level waste would have had a longer cooling time in the pools compared to the minimum cooling time required for the chosen cask technology. This will provide a conservative, safe approach for Koeberg’s facility. Dry cask storage technology has provided a reliable interim dry storage solution for several countries. Despite uncertainties for long term disposal options, the proposed dry cask storage facility at Koeberg is a suitable interim storage alternative for South Africa to allow continuous operation of the plant. This conclusion is based on the physics and engineering aspects that have been presented in this minor dissertation.

This report is concerned with the question whether Accelerator mass spectrometry (AMS) is asuitable technique for measuring actinide inventory in spent nuclear fuel, and if it is better thanpresent techniques for these measurements. AMS is a kind of Mass spectrometry (MS) and has alot of applications where radio carbon dating is one of the most common. AMS has been used formaking measurements on actinides before but mostly from traces in bioassay that could have beenin contact with weapon plutonium, and in bioassay near enrichment plants and reprocessingplants. It is shown in this report that AMS is more sensitive in low level measurements than thecurrent technique for spent nuclear fuel. ICP-MS is the current technique in use for inventorymeasurements on nuclear fuel at Swedish Nuclear Fuel and Waste Management Company (SKB).ICP-MS is also a kind of MS technique which is well-tried for inventory measurements on spentnuclear fuel. The difference in sensitivity ranges in levels of magnitude depending on whichisotope that is interesting for measurements. The lower detection limits for AMS is about 105-107atoms which makes it possible to use samples from nuclear fuel that is in the order of 10-10-10-16gto achieve the lower detection limit. The recommendation from this report is to make studies ifAMS also is an economical and efficiently suitable technique for future use on the actinideinventory in spent nuclear fuel. / Denna rapport handlar om huruvida Accelerator masspektrometri (AMS) är en lämplig teknik förmätning av aktinidinventeriet i använt kärnbränsle. Rapporten går också igenom om AMS är bättreän nuvarande tekniker för dessa mätningar. AMS är en typ av masspektrometri (MS) och har enmängd användningsområden, kol-14 metoden är en av de vanligaste. AMS har också ofta använtsför att göra mätningar på aktinidinnehåll i biomassa som kan ha varit i kontakt medvapenplutonium, och i närheten av anrikningsanläggningar och upparbetningsanläggningar. Detvisas i rapporten att AMS är en mer känslig metod än de nuvarande teknikerna som används förmätningar på aktinidinventariet i använt kärnbränsle. ICP-MS är den aktuella teknik som användsför mätningar på aktinidinventariet i använt kärnbränsle vid Svenska Kärnbränslehantering AB(SKB). ICP-MS är också en typ av MS teknik. MS är väl beprövad för mätningar av inventariet påanvänt kärnbränsle. Skillnaden i känslighet varierar i flera storleksordningar beroende på vilkenisotop som är intressant för mätningarna. Den lägre detektionsgränsen för AMS är cirka 105-107atomer, vilket gör det möjligt att använda prover från kärnbränsle som är i storleksordningen 10-10-10-16g för att uppnå den lägre detektionsgränsen. Rekommendationen från denna rapport är attgöra undersökningar om AMS också är ekonomiskt lönsam och tillräckligt effektiv teknik förframtida bruk inom mätningar av aktinidinventariet i använt kärnbränsle.

The Multi-Isotope Process Monitor is being developed at Pacific Northwest National Laboratory as a method to verify the process conditions within a nuclear fuel reprocessing facility using the gamma spectra of various process streams. The technique uses multivariate analysis techniques such as principal component analysis and partial least squares regression applied to gamma spectra collected of a process stream in order to classify the contents as belonging to a normal versus off-normal chemistry process. This approach to process monitoring is designed to function automatically, nondestructively, and in near real-time. To extend the Multi-Isotope Process Monitor, an analysis method to char- acterize spent nuclear fuel based on the reactor of origin, either pressurized or boiling water reactor, and burnup of the fuel using nuclide concentrations as input data has been developed. While the Multi-Isotope Process Monitor uses gamma spectra as input data, nuclide activities were used in this work as an initial step before Nuclide composition information was generated using ORIGEN-ARP for different fuel assembly types, initial 235U enrichments, burnup values, and cooling times. This data was used to train, tune, and test several multivariate analysis algorithms in order to compare their performance and identify the technique most suited for the analysis. To perform the classification based on reactor type, four methods were considered: k-nearest neighbors, linear and quadratic discriminant analysis, and support vector machines. Each method was optimized, and its performance on a validation set was used to determine the best method for classifying the fuel reactor class. Partial least squares was used to make burnup predictions. Three models were generated and tested: one trained on all the data, one trained for just pressurized water reactors, and one trained for boiling water reactors. Quadratic discriminant analysis was chosen as the best classifier of reactor class because of its simplicity and its potential to be extended to classify spent nuclear fuel’s fuel assembly type, i.e, more specific classes, using nuclide concentrations as input data. In the case of predicting the burnup of spent fuel using partial least squares, it was determined that making reactor-specific partial least squares models, one trained for pressurized water reactors and one trained for boiling water reactors, performed better than a single, general model that was trained for all light water reactors. Thus, the the classifier, regression algorithm, and all the necessary intermediate data processing steps were combined into a single analysis method and implemented as a Matlab function called “burnup.” This function was used to test the analysis routine on an additional set of data generated in ORIGEN-ARP. This dataset included samples with parameters that were not represented in the development data in order to ascertain the analysis method’s ability to analyze data for which it has not been explicitly trained. The algorithm was able to achieve perfect binary classification of the reactor as being a pressurized or boiling water reactor on the dataset and made burnup predictions with an average error of 0.0297%. / text

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Made available in DSpace on 2014-10-09T12:27:08Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T13:56:18Z (GMT). No. of bitstreams: 0 / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Made available in DSpace on 2014-10-09T12:27:08Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T13:56:18Z (GMT). No. of bitstreams: 0 / O combustível nuclear queimado (CNQ) é retirado do reator nuclear após exaurir sua eficiência de geração de energia. Após ser retirado do reator, esse combustível é temporariamente armazenado em piscinas com água na própria instalação do reator. Durante esse tempo, o calor gerado e os elementos radioativos presentes, de meia-vida média e curta, caem até níveis que permitem retirar o combustível queimado da piscina e enviá-lo para depósitos temporários de via seca. Nessa fase, o material precisa ser armazenado segura e eficazmente de forma que possa ser recuperado em futuro próximo, ou disposto como rejeito radioativo. A quantidade de combustível queimado cresce anualmente e, nos próximos anos, vai aumentar mais ainda por causa da construção de novas instalações de geração de energia de origem nuclear. Nos dias de hoje, o número de instalações novas voltou a atingir os níveis da década de 1970, porque é maior que a quantidade de ações de descomissionamento de instalações antigas. Antes que seja tomada qualquer decisão, seja a de recuperar o combustível remanescente ou considerar o CNQ como rejeito radioativo, o mesmo precisa ser isolado em um dos diferentes tipos de armazenagem existentes no mundo. No presente estudo mostra-se que a armazenagem do CNQ, via seca, em cascos é a opção mais vantajosa. Propõe-se um modelo de casco autóctone para combustível de reatores de potência e de uma instalação de armazenagem para abrigar esses cascos. É um estudo multidisciplinar no qual foi desenvolvida a parte conceitual de engenharia e que poderá ser usada para que o CNQ nacional, retirado dos reatores brasileiros de potência, seja armazenado com segurança por um longo período até que as autoridades brasileiras decidam o local para deposição final. / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

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