Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2015-01-19T10:17:33Z No. of bitstreams: 0 / Made available in DSpace on 2015-01-19T10:17:33Z (GMT). No. of bitstreams: 0 / Dissertação (Mestrado em Tecnologia Nuclear) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2015-01-19T10:17:33Z No. of bitstreams: 0 / Made available in DSpace on 2015-01-19T10:17:33Z (GMT). No. of bitstreams: 0 / Neste trabalho foi avaliada a dose efetiva recebida pelos trabalhadores da Gerência de Rejeitos Radioativos do IPEN devido à inalação de 222Rn nos galpões de armazenamento de rejeitos radioativos tratados e não tratados. As concentrações de 222Rn no interior dos galpões foram determinadas por meio da técnica de detecção passiva com detectores de traços nucleares do estado sólido (SSNTD). O detector utilizado foi o CR-39 inserido em uma câmara de difusão do tipo NRPB. Foram monitorados um total de 12 pontos internos e 1 ponto externo no galpão de rejeitos radioativos tratados G4 e 13 pontos no galpão de rejeitos radioativos não tratados G3, durante um período de 11 meses, entre junho de 2012 e maio de 2013. As concentrações variaram de 0,73 ± 0,08 e 4,55 ± 0,16 kBqm-3 entre os períodos de monitoramento no galpão G4 e entre 0,61 ± 0,07 e 2,94 ± 0,12 kBqm-3 no galpão G3. A dose efetiva devido à inalação de 222Rn no interior dos galpões de rejeitos radioativos foi calculada de acordo com os procedimentos da Comissão Internacional de Proteção Radiológica (ICRP) a partir de um fator de conversão de dose, da concentração média do 222Rn no ar e do tempo de exposição dos indivíduos. Os valores de doses apresentados são uma média das concentrações entre os períodos de monitoramento que variam 15,70 mSva-1 no G4 e de 9,27 mSva-1 no G3, sendo que em um dos períodos obteve-se valores superiores ao estabelecidos pelo órgão regulador (CNEN) e recomendados pela Comissão Internacional de Proteção Radiológica (ICRP) de 20 mSva-1 para indivíduos ocupacionalmente expostos, indicando a necessidade de medidas mitigadoras. Cabe, contudo, informar que foi considerada uma hipótese bastante conservativa de 2000 horas de trabalho no local. / Dissertação (Mestrado em Tecnologia Nuclear) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Yucca Mountain, Nevada is the proposed site for the underground storage of high-level civilian nuclear waste in the United States. The repository must be isolated from the general environment for at least 10,000 years. Ground water and gases are potential carriers of radioactive materials. Fractures and connected pores in the host rock are the major pathways for ground water and gases. Therefore, the mechanical and flow properties of the host rock should be understood and utilized in the design of the underground repository. Samples of Topopah Spring tuff from Yucca Mountain were used in this study. Cylindrical specimens were prepared to perform uniaxial and triaxial "damage" tests where specimens are loaded to a particular stress level to induce damage and fracturing and then unloaded. Mechanisms of microcracking at different fracturing levels have been studied by using both an optical microscope and a Scanning Electron Microscope (SEM). The original rock sample without loading was also observed under the microscopes. Many kinds of defects including pores, preexisting fractures, and soft inclusions were found in the undamaged Apache Leap tuff samples. Pores were determined to be the main microstructures in Topopah Spring tuff that could influence the mechanical and hydrologic properties. Under compressive stresses, microcracking initiates from some of the pores. These microcracks will interact and coalescence to form large microcracks or macroscopic cracks as the load is increased. Crack propagation phenomena, such as pore cracking, pore linking, pore collapse and the formation of en echelon arrays were all found in specimens unloaded prior to complete failure. The failure of tuff specimens is often through a process of shear localization. In summary, the deformation and failure of both tuffs occurred by progressive fracturing, starting from microcracking on the small scale, and ending as fractures coalesced to form macroscopic fractures and shear localization. On the basis of the experimental studies, micromechanical models such as the pore collapse model and the pore linking model have been developed based on fracture mechanics theory. These models are used to predict the constitutive behavior for tuff and the predicted stress-strain curves match well with the experimental curves.

This is a summary of findings from an American Indian rapid cultural assessment. As such, this is not a formal report. The text in this summary of findings has been prepared to fit directly into the Intermodal Environmental Assessment (IM EA). This summary of findings was prepared by a study team of Indian people directly from their own field observations. <p> The American Indian Transportation Committee (AITC) was formed in August of 1996 during a study of American Indian issues related to the transportation of Low-level Radioactive Waste to the Nevada Test Site. The AITC contained 9 members who were selected (with the approval of their respective governments) to represent the 29 tribes involved in the study (see Austin 1998:4). The AITC helped with all aspects of that study, they being deeply involved in developing culturally appropriate research methods, helping with the interviews, and closely reviewing the findings. Their efforts were finally presented in a report entitled Native Americans Respond to the Transportation of Low Level Radioactive Waste To the Nevada Test Site (September 1998) edited by Diane Austin.

No description available.

Porous plates delivered calcium chloride at a negative potential to the top of blocks of partially welded (20.1 x 20.1 x 66.6 cm) and densely welded (30.1 x 20.1 x 48.1 cm) tuff with discrete fractures. During infiltration, flux increased through the partially welded block's fracture as the applied suction was lowered to 2.3 cm. The wetting front advanced 66.6 cm in 239 days. Chloride concentration and temporal moments from five tracer tests with 0 to 5 cm of applied suction indicated that preferential fracture flow occurred. Displacement transducer data reflect a decrease in fracture aperture at several months prior to but not during tracer tests. Fracture transmissivities decreased an order of magnitude (6.4 x 10⁻⁹ to 4.2 x 10⁻¹⁰ M²/s) as the applied suction increased from 0 to 5 cm while the tensiometer data indicated a suction of about 20 cm of water within the fracture and matrix. Highest during infiltration to an initially dry block, inflow losses of 3 to 44 percent due to evaporation are the greatest source of error for the constant potential method used.

No description available.

No description available.

Tématem této práce je České atomové právo. Atomové právo je tvořeno soustavou právních norem, které regulují činnosti právnických a fyzických osob související s mírovým využíváním atomové energie a ionizujícího záření. Atomové právo je velmi úzce spojeno s přírodovědnými vědními obory, a proto je pro jeho pochopení důležitá alespoň základní věcná znalost této materie. Stručný souhrn nejdůležitějších vědně-historických milníků uvádím v první kapitole této práce. Následují kapitoly věnované historickému vývoji právní úpravy atomového práva a současné právní úpravě atomového práva. Vývoj atomového práva je nahlížen jak z hlediska mezinárodního a evropského tak národního. Českou právní úpravu atomového práva porovnávám s právní úpravou finskou a zamýšlím se nad jejich odlišnostmi a možnostmi, jak by jedna mohla obohatit druhou. U obou úprav porovnávám kromě formální stránky také nástroje ochrany životního prostředí a zásady, které jsou pro ně stěžejní. Vzhledem k možnému zneužití jaderných materiálů k válečným účelům a rovněž s ohledem na potenciální přeshraniční dopady případné jaderné havárie se právní regulací mírového využívání jaderné energie a ionizujícího záření zabývají mnohé mezinárodní organizace, kterým v této práci pro jejich důležitost a autoritu věnuji samostatnou kapitolu. České atomové...

Técnicas radioquímicas de separação têm sido comumente utilizadas para a caracterização de rejeitos radioativos. No entanto, a determinação de radionuclídeos emissores alfa, beta e gama por análise radioquímica, em amostras de filtro cartucho utilizado no processo de retratamento da água de um reator do tipo piscina, não foi abordada anteriormente na literatura. Este trabalho tem como objetivo estabelecer um método de solubilização para os filtros, identificar e quantificar os radionuclídeos presentes nestes rejeitos, sendo eles os Radionuclídeos Chaves (RC) (60Co, 108mAg, 110mAg) e principalmente os Radionuclídeos de Difícil Medição (RDM) (63Ni, 90Sr, 234U, 235U, 238U, 238Pu, 239+240Pu, 241Pu, 241Am, 242Cm e 243+244Cm), para posterior aplicação do método de Fator de Escala (FE) no trabalho rotineiro de caracterização de rejeitos radioativos. A distribuição dos radionuclídeos no filtro cartucho foi investigada pela determinação dos radionuclídeos emissores gama e os resultados obtidos foram utilizados para calcular o escore-Z. Os resultados indicaram que todos os filtros podem ser considerados homogêneos, atendendo aos critérios de homogeneidade recomendados pela Agência Internacional de Energia Atômica (AIEA), e com isso foi possível definir a quantidade de amostras representativas a serem analisadas. Foram determinados, a partir de dados analíticos, a correlação existente entre os RDMs e RCs selecionados e foram obtidos FEs para todos os RDMs, com exceção do 241Pu. / Separation radiochemical techniques have been commonly used for the characterization of radioactive waste. However, the determination of alpha, beta and gamma emitting radionuclides by radiochemical analysis in cartridge filter samples used in the water re-treatment process of a pool-type reactor was not previously discussed in the literature. This work aims to establish a method of solubilization for the filters, to identify and quantify the radionuclides present in these wastes, being the Key Nuclides (KN) (60Co, 108mAg, 110mAg) and especially the Difficult To Measure radionuclides (DTM) (63Ni , 90Sr, 234U, 235U, 238U, 238Pu, 239+240Pu, 241Pu, 241Am, 242Cm and 243+244Cm), for later application of the Scaling Factors (SF) method in routine radioactive waste characterization work. The distribution of the radionuclides in the cartridge filter was investigated by determination of gamma-emitting radionuclides and the results obtained were used to calculate the Z-score. The results indicated that all the filters can be considered homogeneous, according to the criteria of homogeneity recommended by the International Atomic Energy Agency (IAEA), and with this it was possible to define the quantity of representative samples to be analyzed. From the analytical data, the correlation between the DTMs and the selected KNs was determined and the SFs were obtained for all the DTMs, except for the 241Pu.