Thesis (M.S. - Hydrology and Water Resources)--University of Arizona, 1988. / Includes bibliographical references (leaves 243-246).

Thesis (Ph. D.)--University of Nevada, Reno, 2006. / "August 2006." Includes bibliographical references (leaves 161-169). Online version available on the World Wide Web.

Thesis (Ph. D.)--Florida State University, 2005. / Advisor: Dr. Joel E. Kostka, Florida State University, College of Arts and Sciences, Dept. of Oceanography. Title and description from dissertation home page (viewed June 22, 2005). Document formatted into pages; contains xii, 94 pages. Includes bibliographical references.

Some of the important technical issues involved in the implementation of a spent fuel storage regime under international auspices are discussed. In particular, we consider: the state of the art as far as the different possible storage modes are concerned, the relevant accident, sabotage, and transportation considerations, and the impact of recent technical spent fuel safeguards initiatives on the nonproliferation rationale for international spent fuel management. / Prepared for the U.S. Dept. of Energy under Contract no. EN-77-S-02-4571.A000.

Approximately 96,000 tonnes of the UK Higher Activity Waste (HAW) inventory consists of irradiated nuclear graphite. The current Nuclear Decommissioning Authority (NDA) baseline strategy for irradiated graphite in England and Wales is isolation in a future Geological Disposal Facility, with Scottish policy endorsing an alternative decision of near surface long-term storage. Irradiated graphite disposal routes in the UK remain under review, however, as there are concerns surrounding timing and whether deep geological disposal is the most appropriate course of action for graphite. An alternative waste management solution is treatment prior to disposal to separate mobile radioactive isotopes such as 3H and 14C from the bulk material, allowing for HAW volume reduction and concentration. Optimisation of an existing thermal treatment process at the Nuclear Graphite Research Group (NGRG) of the University of Manchester has been effected and a detailed review of the uncertainties associated with quantitative determination of radioisotope releases during thermal treatment of irradiated graphite samples has been conducted. Thermal treatment experiments in both an inert atmosphere and 1% oxygen in argon atmosphere have been conducted for temperatures ranging from 600°C to 800°C, and durations from 4 to 120 hours, to determine the effects of oxidation time and temperature, and the consequent oxidation characteristics on the release rate of prominent radioisotopes, with a focus on the release of 14C. Lower temperature treatments in an oxidising atmosphere have shown that a preferential release of 14C-enriched graphite can be achieved from the bulk material of Oldbury Magnox reactor irradiated graphite, with evidence demonstrating that this liberated 14C-enriched region is located at the graphite surfaces throughout the porous structure. A large proportion of radiocarbon found in this irradiated graphite, however, is uniformly distributed throughout the bulk material and cannot be selectively oxidised. It is found that prominent metallic radioisotopes such as 60Co are not mobile at these temperatures and remain in the bulk graphite material, inclusive of radioactive caesium which the literature suggests will volatilise. The preliminary results were undertaken as part of the EU FP7 EURATOM Project: CARBOWASTE.

Made available in DSpace on 2014-10-09T12:45:05Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:07:15Z (GMT). No. of bitstreams: 1 07015.pdf: 5319604 bytes, checksum: 27a37f27f23ef592fabde0745b070b1f (MD5) / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Made available in DSpace on 2014-10-09T12:53:25Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:09:32Z (GMT). No. of bitstreams: 1 12213.pdf: 5005382 bytes, checksum: c4bde457760b3a6d6f53b64c21e33010 (MD5) / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Made available in DSpace on 2014-10-09T12:45:05Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:07:15Z (GMT). No. of bitstreams: 1 07015.pdf: 5319604 bytes, checksum: 27a37f27f23ef592fabde0745b070b1f (MD5) / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Made available in DSpace on 2014-10-09T12:53:25Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:09:32Z (GMT). No. of bitstreams: 1 12213.pdf: 5005382 bytes, checksum: c4bde457760b3a6d6f53b64c21e33010 (MD5) / No presente trabalho, foi utilizada a técnica de aquecimento por microondas para estudar a imobilização de rejeitos radioativos de nível de atividade baixo e médio, como resinas de troca iônica exauridas, empregadas na remoção de íons indesejáveis dos circuitos primários de refrigeração de reatores nucleares refrigerados a água, e aquelas usadas em colunas de separação química e radionuclídica no controle de qualidade de radioisótopos. Matrizes betuminosas reforçadas com alguns tipos de borrachas (Neoprene®, Silicone e Etileno Vinil Acetato - EVA), provenientes de material descartado ou sobras de produção, foram utilizadas para incorporação dos rejeitos radioativos. As irradiações das amostras foram feitas em um forno de microondas caseiro, que opera com freqüência de 2.450MHz e possui potência de 1.000W. As amostras foram caracterizadas, empregando-se ensaios de penetração, resistência à lixiviação, pontos de amolecimento, fulgor e combustão, termogravimetria e microscopia óptica. Os resultados obtidos mostraram-se compatíveis com os padrões dos componentes das matrizes, indicando que esta técnica é uma alternativa bastante útil aos métodos de imobilização convencionais e para esses tipos de rejeitos radioativos. / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Atualmente, a caracterização dos resíduos radioativos gerados na operação do reator nuclear de pesquisas IEA-R1 está em curso. O reator IEA-R1 é um reator do tipo piscina aberta, moderado e refrigerado por água leve, utilizando dois leitos de resinas de troca iônica e de carvão ativado para purificação de água de refrigeração. Estes meios filtrantes são substituídos quando já não são capazes de manter a qualidade da água dentro dos limites exigidos e são tratados como rejeitos radioativos. Contendo produtos de fissão, ativação e actinídeos que escapam do núcleo do reator para a água da piscina, apresentam altas taxas de dose devido à quantidade de emissores gama de meias-vidas curtas e intermediárias, emissores alfa, elementos transurânicos de meia-vida longa bem como emissores beta puros. A caracterização destes rejeitos, consequentemente, requer métodos de análise radioquímica que incluem a amostragem e o processamento das amostras, resultando em doses elevadas para os trabalhadores. Nesse contexto, o objetivo deste trabalho consistiu em correlacionar os resultados das análises radioquímicas de amostras de rejeitos, com os resultados das medições radiométricas, utilizando a modelagem das taxas de dose em diferentes distâncias da superfície dos embalados. As taxas de dose medidas foram comparadas com os resultados de cálculos . Massa, volume e geometria das fases sólidas e líquidas de cada um dos tambores também foram determinadas, uma vez que o teor de água varia amplamente entre diferentes tambores, e são essenciais para estimar as atividades totais em cada tambor. / The characterization of the radioactive wastes generated in the operation of the nuclear research reactor IEA-R1 is currently ongoing. The IEA-R1 is an open pool type reactor, moderated and cooled by light water that uses two beds of ion-exchange resins and activated charcoal to remove impurities from the cooling water. These filter media are replaced when they are no longer able to maintain water quality within the required limits and are treated as radioactive waste. They contain the actinides and the fission and activation products that leaked into the reactor pool water. They give off high dose rates due to the amount of gamma-emitters present and are a long-term radiation safety concern because of their content of long-lived alpha- and beta-emitters. The characterization of these wastes requires radiochemical analysis methods, which include the sampling and processing of samples, resulting in high exposure to the workers. The objective of this study was to correlate the results of activity concentrations obtained in previous radiochemical analyses with the results of measurements of dose rates at various distances from the package surfaces, aiming at reducing the exposure of personnel by avoiding more sampling and sample analysis operations. Mass, volume and geometry of solid and liquid phases of each drum, which vary widely among different drums, were also estimated and use to determine total activity. The measured and calculated dose rates were compared to confirm the activity estimates.