Radioactivity is an important part of the environment. In the years 2014 - 2015 a survey on medicinal radioactive springs was carried out in the Tanvald granite body. During extensive radiohydrogeochemical exploration in the surrounding area of Schindler spring, the outcrop of U - mineralization has been found at a new construction site in the Kokonín fault. It is possible to find "hot" pieces of uranium ore fragments in the quaternary cover in this place. The occurrence of the hydrothermal (vein) type of uranium in the Krkonoše - Jizera is new as it hasn't been mentioned in literatures before. After an agreement with the owners, detailed radiometric survey proceeded in the surrounding area of two affected houses by using gamma spectrometry and emanometry at the dense net. There is a wider area between two houses with activities of 222 Rn over 1 MBq/m3 and on the ore outcrop was measured 3.3 MBq/m3 . A quaternary solifluction flow with an increased radioactivity creeps down under the House 1. The highest uranium contents reached up to 291 ppm eU (= 3 595 Bq/kg 226 Ra) on the uranium ore lens outcrop. The average of gamma dose rate is 85,6 nGy/h with maximem 261 nGy/h on the outcrop. A further radiometric survey through the entire Kokonín fault line was carried out in an area of roughly 0.25 x 1 km2...

The neutron plays a central role in numerous fields of physics, a fact that entails a need for methods of measuring neutron energy spectra. In this project, a technique for neutron spectrometry through measurements of neutron-induced radioactivity in activation detectors was developed and tested. The developed technique involves irradiating element samples with neutrons, measuring activation products with a gamma spectrometer, and then performing a neutron spectrum unfolding procedure. The elements indium, iron, magnesium, aluminium, zinc, titanium, and copper were used as activation detectors and irradiated with neutrons from an americium-beryllium (AmBe) neutron source. Subsequent gamma spectrometry was performed with the UGGLA high-purity germanium detector setup at Uppsala University. The GRAVEL unfolding algorithm was implemented in MATLAB and used to unfold neutron spectra based on an initial spectrum guess. The unfolded neutron spectrum agrees well with the expected AmBe spectrum, though some difference between the spectra is attributed to neutron scattering in the irradiation environment. A possible ability to find approximate neutron spectra from inaccurate initial guesses is found, but additional work is needed to understand better how the initial guess affects the result for different neutron sources. Because activation detectors do not require electrical power when measuring neutrons, can be made sensitive to a wide range of neutron energies, and do not detect other types of radiation, future applications could find the developed neutron spectrometry method practical.

Os experimentos realizados e apresentados nesta dissertação resultaram na caracterização neutrônica de núcleos na configuração cilíndrica com 30 varetas combustíveis de diâmetro com um espaço, criado pela retirada de 16 varetas centrais, preenchido com água leve (H2O) ou água pesada (D2O) no reator nuclear de pesquisa IPEN/MB-01. Nestes núcleos, efetuou-se experimentos de correlação de canais nucleares, calibração de barras de controle e irradiação de detectores de ativação de diversos materiais em forma de folhas, cujas faixas energéticas de atuação abrangem grande parte do espectro de nêutrons do núcleo do reator, para a obtenção de parâmetros nucleares, como excessos de reatividade, reatividades totais, atividades saturadas por núcleo alvo, razões espectrais, razões de cádmio e fluxo de nêutrons multigrupo. Com a irradiação de fios de ativação de ouro na parte radial foram obtidas as formas espaciais dos fluxos de nêutrons térmicos e epitérmicos. Os resultados mostraram as características espectrais dessa nova configuração com o espaço das 16 varetas combustíveis preenchido com os dois materiais moderadores. No espaço com a água leve houve um aumento significativo de 294% do fluxo de nêutrons térmicos em comparação com a configuração padrão retangular de 28×26 varetas combustíveis. Com a água pesada aumentou-se a reatividade do sistema com ρ = (783 ± 54) pcm a mais de excesso de reatividade que na configuração com água leve. Os resultados calculados foram simulados nos códigos computacionais MCNP5, SANDBP e CITATION, onde se obtiveram resultados acurados e precisos para as atividades saturadas por núcleo alvo, as distribuições energéticas e espaciais dos fluxos de nêutrons da parte ativa e de parte do refletor e as comparações diretas das seções de choque entre as razões espectrais experimentais e calculadas. / The experiments performed and presented in this thesis results in the neutronic characterization of the core with cylindrical configuration with 30 fuel rods diameter and a space, created by the removal of 16 central rods, filled with light water (H2O) or heavy water (D2O) in the IPEN/MB-01 nuclear research reactor. In these cores were performed experiments of nuclear channels correlation, control rod worth and irradiation of activation detectors of various materials in foils shapes, whose energy performances cover much of the reactor core neutron spectrum, to obtain nuclear parameters, such as, reactivity excesses, total reactivities, saturated activities per target nucleus, spectral ratios, cadmium ratios and multigroup neutron flux. Activation gold wires detectors were irradiated in radial part to obtain the spatial forms of thermal and epithermal neutron fluxes. The results show the spectral characteristics of this new configuration with the space of 16 fuel rods filled with the two moderator materials. In the space with light water there was a significant increase of 294% of the thermal neutron flux compared to standard rectangular configuration of 28×26 fuel rods. With heavy water the system reactivity was increased, more ρ = (783 ± 54) pcm in excess reactivity than in the light water configuration. The calculated results were simulated in computational codes MCNP5, SANDBP and CITATION, where accurate and precise results were obtained for saturated activities per target nucleus, the energy and spatial distributions of the neutron fluxes for the active part and part of the reflector and the direct comparisons of cross sections between the experimental and calculated spectral ratios.

O 99mTc é o radionuclídeo mais utilizado em medicina nuclear. No Brasil os geradores de 99Mo/99mTc são produzidos exclusivamente pelo Centro de Radiofarmácia do IPEN-CNEN/SP, com 99Mo importado de diferentes fornecedores. O 99Mo (t1/2 = 66 h), por ser um produto de fissão do 235U, pode conter impurezas radionuclídicas prejudiciais à saúde humana. Dessa forma, para que o gerador seja utilizado de forma segura, é necessário que o 99Mo seja avaliado por ensaios de controle de qualidade e atenda à alguma especificação descrita em farmacopeia. A Farmacopeia Europeia (FE) apresenta monografia, com parâmetros (identificação, pureza radioquímica e pureza radionuclídica), métodos de análise, e limites, para avaliação da qualidade da solução de [99Mo] na forma de molibdato de sódio, que é utilizada como matéria-prima no preparo dos geradores de 99Mo/99mTc. No entanto, observa-se uma dificuldade na implementação e execução dos métodos por parte dos produtores de geradores, com pouca literatura sobre o assunto, provavelmente devido à falta de praticidade dos métodos propostos e à extensa lista de reagentes utilizados. Nesse trabalho foram avaliados vários parâmetros de qualidade do 99Mo descritos na monografia da FE. Foram estudados métodos de separação do 99Mo de suas impurezas radionuclídicas por extração em fase sólida (SPE) e por TLC. Após separação por SPE, foi proposta a quantificação de metais por ICP-OES para avaliar a porcentagem de retenção de Mo e a porcentagem de recuperação de Ru e Te e Sr em diversos tipos de cartuchos, em substituição ao uso de radiotraçadores. Observou-se que a marca de cartucho de SPE para separação do 99Mo recomendada pela FE apresentou baixa recuperação para Ru, quando comparado aos outros cartuchos de troca aniônica disponíveis no mercado. Amostras de 99Mo de diferentes fornecedores mundiais foram analisadas. Observou-se que é possível realizar a quantificação de 103Ru em amostras de 99Mo mesmo com tempos de decaimento acima de 4 semanas. Um método alternativo de separação do 99Mo do 131I por TLC apresentou resultados promissores. Não foi feita a quantificação das impurezas radionuclídicas emissoras beta e alfa. Todas as amostras analisadas apresentaram resultados dentro das especificações da FE para pureza radioquímica (>95%) e pureza radionuclídica. / 99mTc is the most used radionuclide in nuclear medicine. In Brazil, the 99Mo/99mTc generators are exclusively produced by Radiopharmacy Center at IPENCNEN/ SP, by importing 99Mo from different suppliers. 99Mo (t1/2 = 66 h) is a fission product of 235U and it can have radionuclidic impurities that are prejudicial for human health. For safe use of generators, it is necessary to perform the evaluation of 99Mo by quality control tests in order to assess if 99Mo complies with the specifications. The European Pharmacopoeia (EP) presents a monograph for evaluation of the quality of the [99Mo] solution as sodium molybdate,that is used as raw material for 99Mo/99mTc generators production, including specification parameters (identification, radiochemical purity and radionuclidic purity), analysis methods and limits. However, it has been observed difficulties on the execution and implementation of these methods by the generators producers, with a few literature about this subject, probably due to complexity of the proposed methods. In this work, many quality control parameters of 99Mo described in the EP monograph were evaluated. Separation methods for 99Mo from its radionuclidic impurities by solid phase extraction (SPE) and TLC were studied. After SPE separation, the quantification of metals by ICP-OES to evaluate the percentage of retention of Mo and the percentage of recovery of Ru, Te and Sr using different types of cartridges were proposed, replacing radiotracers use. It was observed that the specific type of SPE cartridge recommended by the EP for separation of 99Mo presented low recoveries for Ru, compared to other available anion exchange SPE cartridges. 99Mo samples from different worldwide suppliers were analyzed. It was observed that quantification of 103Ru in 99Mo samples with decay time higher than 4 weeks is possible. An alternative method for separation of 131I from 99Mo showed promising results by TLC. The quantification of beta and alpha emitters radionuclidic impurities was not performed. All analysed samples presented results that comply with EP specifications for radiochemical purity (>95%) and for radionuclidic purity.

Os experimentos realizados e apresentados nesta dissertação resultaram na caracterização neutrônica de núcleos na configuração cilíndrica com 30 varetas combustíveis de diâmetro com um espaço, criado pela retirada de 16 varetas centrais, preenchido com água leve (H2O) ou água pesada (D2O) no reator nuclear de pesquisa IPEN/MB-01. Nestes núcleos, efetuou-se experimentos de correlação de canais nucleares, calibração de barras de controle e irradiação de detectores de ativação de diversos materiais em forma de folhas, cujas faixas energéticas de atuação abrangem grande parte do espectro de nêutrons do núcleo do reator, para a obtenção de parâmetros nucleares, como excessos de reatividade, reatividades totais, atividades saturadas por núcleo alvo, razões espectrais, razões de cádmio e fluxo de nêutrons multigrupo. Com a irradiação de fios de ativação de ouro na parte radial foram obtidas as formas espaciais dos fluxos de nêutrons térmicos e epitérmicos. Os resultados mostraram as características espectrais dessa nova configuração com o espaço das 16 varetas combustíveis preenchido com os dois materiais moderadores. No espaço com a água leve houve um aumento significativo de 294% do fluxo de nêutrons térmicos em comparação com a configuração padrão retangular de 28×26 varetas combustíveis. Com a água pesada aumentou-se a reatividade do sistema com ρ = (783 ± 54) pcm a mais de excesso de reatividade que na configuração com água leve. Os resultados calculados foram simulados nos códigos computacionais MCNP5, SANDBP e CITATION, onde se obtiveram resultados acurados e precisos para as atividades saturadas por núcleo alvo, as distribuições energéticas e espaciais dos fluxos de nêutrons da parte ativa e de parte do refletor e as comparações diretas das seções de choque entre as razões espectrais experimentais e calculadas. / The experiments performed and presented in this thesis results in the neutronic characterization of the core with cylindrical configuration with 30 fuel rods diameter and a space, created by the removal of 16 central rods, filled with light water (H2O) or heavy water (D2O) in the IPEN/MB-01 nuclear research reactor. In these cores were performed experiments of nuclear channels correlation, control rod worth and irradiation of activation detectors of various materials in foils shapes, whose energy performances cover much of the reactor core neutron spectrum, to obtain nuclear parameters, such as, reactivity excesses, total reactivities, saturated activities per target nucleus, spectral ratios, cadmium ratios and multigroup neutron flux. Activation gold wires detectors were irradiated in radial part to obtain the spatial forms of thermal and epithermal neutron fluxes. The results show the spectral characteristics of this new configuration with the space of 16 fuel rods filled with the two moderator materials. In the space with light water there was a significant increase of 294% of the thermal neutron flux compared to standard rectangular configuration of 28×26 fuel rods. With heavy water the system reactivity was increased, more ρ = (783 ± 54) pcm in excess reactivity than in the light water configuration. The calculated results were simulated in computational codes MCNP5, SANDBP and CITATION, where accurate and precise results were obtained for saturated activities per target nucleus, the energy and spatial distributions of the neutron fluxes for the active part and part of the reflector and the direct comparisons of cross sections between the experimental and calculated spectral ratios.

Depuis le début des années 1980, le Commissariat à l’Énergie Atomique développe et met en oeuvre un système de spectrométrie gamma aéroportée, appelé HELINUCTM. Ce système, composé de détecteurs NaI(Tl) d’un volume de 16 L, est utilisé afin d’établir un état des lieux radiologique des sites survolés. Les principales missions du système HELINUC consistent en la réalisation de contrôles environnementaux, l’intervention en situation de crise et la recherche de sources ponctuelles. La réalisation de ces missions nécessite le développement de méthodes d’analyse adaptées. L’approche considérée dans cette thèse repose sur une rupture conceptuelle de l’analyse des spectres par rapport aux méthodes utilisées jusqu’alors au CEA : l’analyse ne repose plus sur une considération individuelle et séquentielle des mesures aéroportées, mais sur une considération globale et simultanée de celles-ci. L’étude et le développement de méthodes statistiques adaptées à la quantification des radionucléides naturels et du 137Cs (de 600 keV à 3 MeV), à l’estimation de la contamination en 241Am (basse énergie, inférieure à100 keV) en cas de crise radiologique et la détection de sources ponctuelles (moyenne énergie, entre 100 keV et600 keV) permettent d’améliorer la précision sur les activités déterminées en vol et la détection de sources de faibles activités. / Since the beginning of 1980’s, the CEA has been developing an airborne gamma spectrometry (AGS) system called HELINUCTM using large volume (16 L) NaI(Tl) detectors. HELINUC is used to produce radioactivity mapping of the soil. The different missions of HELINUC are environmental control of radioactivity, nuclear emergency response and research of orphan sources. The continuous development of analysis methods is then required.The approach considered in this thesis is based on a conceptual break from the analysis of spectra compared to the methods used at the CEA until now: the analysis does not rely on an individual and sequential consideration of airborne measurements, but on an overall and simultaneous consideration of them. The study and development of statistical methods for the quantification of natural radionuclides and 137Cs (from 600 keV to 3 MeV), for the estimation of 241Am contamination (low energy, inferior to 100 keV) in case of radiological emergency and for the detection of orphan sources (medium energy, between 100 keV and 600 keV) improve the accuracy of activities estimation and detection of low activities sources.

O 99mTc é o radionuclídeo mais utilizado em medicina nuclear. No Brasil os geradores de 99Mo/99mTc são produzidos exclusivamente pelo Centro de Radiofarmácia do IPEN-CNEN/SP, com 99Mo importado de diferentes fornecedores. O 99Mo (t1/2 = 66 h), por ser um produto de fissão do 235U, pode conter impurezas radionuclídicas prejudiciais à saúde humana. Dessa forma, para que o gerador seja utilizado de forma segura, é necessário que o 99Mo seja avaliado por ensaios de controle de qualidade e atenda à alguma especificação descrita em farmacopeia. A Farmacopeia Europeia (FE) apresenta monografia, com parâmetros (identificação, pureza radioquímica e pureza radionuclídica), métodos de análise, e limites, para avaliação da qualidade da solução de [99Mo] na forma de molibdato de sódio, que é utilizada como matéria-prima no preparo dos geradores de 99Mo/99mTc. No entanto, observa-se uma dificuldade na implementação e execução dos métodos por parte dos produtores de geradores, com pouca literatura sobre o assunto, provavelmente devido à falta de praticidade dos métodos propostos e à extensa lista de reagentes utilizados. Nesse trabalho foram avaliados vários parâmetros de qualidade do 99Mo descritos na monografia da FE. Foram estudados métodos de separação do 99Mo de suas impurezas radionuclídicas por extração em fase sólida (SPE) e por TLC. Após separação por SPE, foi proposta a quantificação de metais por ICP-OES para avaliar a porcentagem de retenção de Mo e a porcentagem de recuperação de Ru e Te e Sr em diversos tipos de cartuchos, em substituição ao uso de radiotraçadores. Observou-se que a marca de cartucho de SPE para separação do 99Mo recomendada pela FE apresentou baixa recuperação para Ru, quando comparado aos outros cartuchos de troca aniônica disponíveis no mercado. Amostras de 99Mo de diferentes fornecedores mundiais foram analisadas. Observou-se que é possível realizar a quantificação de 103Ru em amostras de 99Mo mesmo com tempos de decaimento acima de 4 semanas. Um método alternativo de separação do 99Mo do 131I por TLC apresentou resultados promissores. Não foi feita a quantificação das impurezas radionuclídicas emissoras beta e alfa. Todas as amostras analisadas apresentaram resultados dentro das especificações da FE para pureza radioquímica (>95%) e pureza radionuclídica. / 99mTc is the most used radionuclide in nuclear medicine. In Brazil, the 99Mo/99mTc generators are exclusively produced by Radiopharmacy Center at IPENCNEN/ SP, by importing 99Mo from different suppliers. 99Mo (t1/2 = 66 h) is a fission product of 235U and it can have radionuclidic impurities that are prejudicial for human health. For safe use of generators, it is necessary to perform the evaluation of 99Mo by quality control tests in order to assess if 99Mo complies with the specifications. The European Pharmacopoeia (EP) presents a monograph for evaluation of the quality of the [99Mo] solution as sodium molybdate,that is used as raw material for 99Mo/99mTc generators production, including specification parameters (identification, radiochemical purity and radionuclidic purity), analysis methods and limits. However, it has been observed difficulties on the execution and implementation of these methods by the generators producers, with a few literature about this subject, probably due to complexity of the proposed methods. In this work, many quality control parameters of 99Mo described in the EP monograph were evaluated. Separation methods for 99Mo from its radionuclidic impurities by solid phase extraction (SPE) and TLC were studied. After SPE separation, the quantification of metals by ICP-OES to evaluate the percentage of retention of Mo and the percentage of recovery of Ru, Te and Sr using different types of cartridges were proposed, replacing radiotracers use. It was observed that the specific type of SPE cartridge recommended by the EP for separation of 99Mo presented low recoveries for Ru, compared to other available anion exchange SPE cartridges. 99Mo samples from different worldwide suppliers were analyzed. It was observed that quantification of 103Ru in 99Mo samples with decay time higher than 4 weeks is possible. An alternative method for separation of 131I from 99Mo showed promising results by TLC. The quantification of beta and alpha emitters radionuclidic impurities was not performed. All analysed samples presented results that comply with EP specifications for radiochemical purity (>95%) and for radionuclidic purity.

La dosimétrie neutronique en réacteur se base sur l'analyse de l'activité de dosimètres irradiés, dont certains isotopes-cibles sont l'objet de réactions d'activation ou de fission sous l'effet des neutrons. Les différentes cibles sont sensibles aux neutrons d’énergie particulière. La caractérisation des spectres neutroniques est bien établie dans les domaines thermique, epithermique (Eneutron <1 keV) et rapide (Eneutron >1 MeV), mais il y a une absence de détecteur dans le domaine énergétique entre 1 keV et 1 MeV. Le travail de thèse a abouti sur un choix final: la capture (n, &#947;) sur les isotopes 92Zr et 94Zr, présents dans le zirconium naturel, pour former les isotopes 93Zr (stable) et 95Zr (radioactif). L'expérience ZIMA a été réalisée sur le réacteur OSIRIS pour démontrer la faisabilité de la méthode de détection proposée. Les analyses post-irradiation sont la spectrométrie &#947; et la spectrométrie de masse par accélérateur. Pour analyser les résultats expérimentaux, ZIMA a été simulée avec le code neutronique TRIPOLI-4 basé sur la méthode de Monte Carlo. Les rapports Calcul/Expérience présentés dans la thèse permettent de conclure que la détection neutronique (1 keV – 1 MeV) par capture de 94Zr et 92Zr donne des résultats probants. Les mesures obtenues sont exploitables. / Reactor dosimetry goal is to reconstruct neutron spectrum in a particular reactor location. Today we can reconstruct with precision thermal (<eV) and fast (>MeV) parts of neutron spectrum by using dosimeters with an adequate sensitivity. Nowadays there is no dosimeter for the intermediate energy region 1 keV - 1 MeV. Thus, the PhD goal is to select the 1 keV - 1 MeV sensible target-isotope and nuclear reaction and verify our solution by experimental irradiation. PhD final choice is for neutron capture reaction (n, &#947;) on 92Zr and 94Zr. Neutron irradiation produces 2 isotopes: 93Zr and 95Zr, stable and radioactive. Irradiation experiment was performed in OSIRIS reactor. Post-irradiation analyses of irradiated Zr samples are &#947; spectrometry and Accelerator Mass Spectrometry. In order to simulate irradiation experiment we performed calculation with neutron transport code TRIPOLI-4, based on Monte Carlo method. The goal of ZIMA (Zirconium Irradiation for Mass and Activity analysis) experiment was to prove the feasibility of 1 keV - 1 MeV neutron detection by (n,&#947;) capture on 92Zr and 94Zr under boron nitride filter. C/E ratios presented in this PhD allow us to conclude that activation of 94Zr and 92Zr gives us acceptable results.

Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2015-01-19T10:26:11Z No. of bitstreams: 0 / Made available in DSpace on 2015-01-19T10:26:11Z (GMT). No. of bitstreams: 0 / Dissertação (Mestrado em Tecnologia Nuclear) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Jusqu’au début des années 1980, la spectrométrie gamma aéroportée a avant tout été utilisée pour des applications géophysiques et ne concernait que la mesure des concentrations dans les sols des trois radionucléides naturels (K40, U238 et Th232). Durant les quinze dernières années, un grand nombre de dispositifs de mesures a été développé, la plupart après l’accident de Tchernobyl, pour intervenir en cas d’incidents nucléaires ou de surveillance de l’environnement. Les algorithmes développés ont suivi les différentes missions de ces systèmes. La plupart sont dédiés à l’extraction des signaux à moyenne et haute énergie, où les radionucléides naturels (K40, les chaînes U238 et Th232) et les produits de fission (Cs137 et Co60 principalement) sont présents. A plus basse énergie (< 400 keV), ces méthodes peuvent toujours être utilisées mais les particularités du fond de diffusion, très intense, les rendent peu précises. Cette zone énergétique est importante : les SNM émettent à ces énergies. Un algorithme, appelé 2-fenêtres (étendu à 3), a été développé permettant une extraction précise et tenant compte des conditions de vol. La surveillance du trafic de matières radioactives dans le cadre de la sécurité globale a fait son apparition depuis quelques années. Cette utilisation nécessite non plus des méthodes sensibles à un élément particulier mais des critères d’anomalie prenant en compte l’ensemble du spectre enregistré. Il faut être sensible à la fois aux radionucléides médicaux, industriels et nucléaires. Ce travail a permis d’identifier deux familles d’algorithmes permettant de telles utilisations. Enfin, les anomalies détectées doivent être identifiées. La liste des radionucléides nécessitant une surveillance particulière, recommandée par l’AIEA, contient une trentaine d’émetteurs. Un nouvel algorithme d’identification a été entièrement développé, permettant de s’appuyer sur plusieurs raies d’absorption par élément et de lever les conflits d’identification. / Airborne gamma spectrometry was first used for mining prospection. Three main families were looked for: K40, U238 and Th232. The Chernobyl accident acted as a trigger and for the last fifteen years, a lot of new systems have been developed for intervention in case of nuclear accident or environmental purposes. Depending on their uses, new algorithms were developed, mainly for medium or high energy signal extraction. These spectral regions are characteristics of natural emissions (K40, U238- and Th-232 decay chains) and fissions products (mainly Cs137 and Co60). Below 400 keV, where special nuclear materials emit, these methods can still be used but are greatly imprecise. A new algorithm called 2-windows (extended to 3), was developed. It allows an accurate extraction, taking the flight altitude into account to minimize false detection. Watching radioactive materials traffic appeared with homeland security policy a few years ago. This particular use of dedicated sensors require a new type of algorithms. Before, one algorithm was very efficient for a particular nuclide or spectral region. Now, we need algorithm able to detect an anomaly wherever it is and whatever it is : industrial, medical or SNM. This work identified two families of methods working under these circumstances. Finally, anomalies have to be identified. IAEA recommend to watch around 30 radionuclides. A brand new identification algorithm was developed, using several rays per element and avoiding identifications conflicts.