In dieser prospektiven, deskriptiven Querschnitts- und Pilotstudie sollten die Radioisotopen-Aufnahmemuster (radioisotope uptake - RU) der Reservekrone und des parodontalen Knochens der Ober- und Unterkieferbackenzähne (CT) bei klinisch gesunden Pferden beschrieben und die Auswirkungen des Alters auf die RU bewertet werden.:Table of Contents Abbreviations: .......................................................................................................... VI 1. Introduction ........................................................................................................ 1 2. Literature overview ............................................................................................ 3 2.1. Evolution of equine dentistry ......................................................................... 3 2.2. Epidemiology of equine dental pathology ..................................................... 5 2.3. Diagnostic imaging modality and equine dental disorders ............................ 5 2.4. Bone scintigraphy as diagnostic tool of equine dental disorders .................. 6 2.5. Literature review of equine dental scintigraphy ............................................ 8 3. Publication ........................................................................................................ 10 Scintigraphic evaluation of the cheek teeth in clinically sound horses ............ 10 3.1. Author contributions .................................................................................... 11 3.2. Abstract ....................................................................................................... 12 3.3. Introduction ................................................................................................. 12 3.4. Material and methods ................................................................................. 14 3.4.1. Subject selection ...................................................................................... 14 3.4.2. Scintigraphic examination ........................................................................ 14 3.4.3. Pilot study ................................................................................................ 15 3.4.4. Image processing and analysis ................................................................ 16 3.4.5. Statistical analysis .................................................................................... 16 3.5. Results ........................................................................................................ 17 3.6. Discussion .................................................................................................. 18 3.7. References ................................................................................................. 22 4. Discussion ........................................................................................................ 31 4.1. Animals ....................................................................................................... 31 4.2. Methodology ............................................................................................... 31 4.3. Results ........................................................................................................ 33 4.4. Study limitation ........................................................................................... 38 4.5. Clinical relevance ........................................................................................ 38 5. Zusammenfassung .......................................................................................... 40 6. Summary ........................................................................................................... 42 7. References ........................................................................................................ 44 8. Acknowledgements ......................................................................................... 51 / This prospective, cross-sectional, descriptive and pilot-designed study aimed to describe the radioisotope uptake (RU) patterns of the reserved crown and periodontal bone of the maxillary and mandibular cheek teeth (CT) in clinically sound horses and to evaluate the age effect on RU. For this purpose, 60 horses that underwent a bone scintigraphy for reason unrelated to head were included and divided equally into four age groups.:Table of Contents Abbreviations: .......................................................................................................... VI 1. Introduction ........................................................................................................ 1 2. Literature overview ............................................................................................ 3 2.1. Evolution of equine dentistry ......................................................................... 3 2.2. Epidemiology of equine dental pathology ..................................................... 5 2.3. Diagnostic imaging modality and equine dental disorders ............................ 5 2.4. Bone scintigraphy as diagnostic tool of equine dental disorders .................. 6 2.5. Literature review of equine dental scintigraphy ............................................ 8 3. Publication ........................................................................................................ 10 Scintigraphic evaluation of the cheek teeth in clinically sound horses ............ 10 3.1. Author contributions .................................................................................... 11 3.2. Abstract ....................................................................................................... 12 3.3. Introduction ................................................................................................. 12 3.4. Material and methods ................................................................................. 14 3.4.1. Subject selection ...................................................................................... 14 3.4.2. Scintigraphic examination ........................................................................ 14 3.4.3. Pilot study ................................................................................................ 15 3.4.4. Image processing and analysis ................................................................ 16 3.4.5. Statistical analysis .................................................................................... 16 3.5. Results ........................................................................................................ 17 3.6. Discussion .................................................................................................. 18 3.7. References ................................................................................................. 22 4. Discussion ........................................................................................................ 31 4.1. Animals ....................................................................................................... 31 4.2. Methodology ............................................................................................... 31 4.3. Results ........................................................................................................ 33 4.4. Study limitation ........................................................................................... 38 4.5. Clinical relevance ........................................................................................ 38 5. Zusammenfassung .......................................................................................... 40 6. Summary ........................................................................................................... 42 7. References ........................................................................................................ 44 8. Acknowledgements ......................................................................................... 51

In the field of nuclear medicine, radioisotopes are used for applications such as diagnostic imag- ing, treatment, and equipment sterilization. The most commonly used radioisotope in medicine is technetium-99m (Tc-99m). It is used in 80% of all nuclear medicine procedures. Its parent isotope is molybdenum-99 (Mo-99). NRU, which is now closed, formerly produced 40% of the worlds demand for Mo-99. The production capacity of this reactor has been supplemented by a network of cyclotrons and a modified research reactor. This study aims to provide an alternative means of production for Mo-99, as well as other radioisotopes by modifying the center pin of a standard 37-element bundle of a CANDU reactor. The neutron transport code DRAGON, and the neutron diffusion code DONJON were used to model a CANDU-9 reactor. The lowest, median, and highest power channels were chosen as candi- dates for the modified bundles. It was found that the reactor parameters were altered by a negligible amount when any one channel was used to house the modified bundles. Significant quantities of the radioisotope lutetium-177 as well as the generating isotopes of the alpha-emitting radioisotopes lead- 212/bismuth-212, and radium-223 were produced. However, only minute amounts of molybdenum-99, and the generating isotope of bismuth-213 were produced. / Thesis / Master of Applied Science (MASc)

INTRODUÇÃO: Os rins são alvos de diversas patologias que comprometem o seu funcionamento em graus variáveis, e em muitos casos, a nefrectomia é o melhor tratamento disponível. Atualmente existe um grande interesse na literatura no intuito de tentar prever como será a evolução da função renal do paciente após uma nefrectomia. Alguns estudos retrospectivos propuseram métodos para esta estimativa, no entanto, suas metodologias possuem falhas e seus resultados são contraditórios. OBJETIVOS: Avaliar de maneira prospectiva os fatores prognósticos préoperatórios associados a evolução da função renal seis meses após nefrectomias unilaterais utilizando para esta análise, um método radioisotópico de referência (51Cr- EDTA). Em seguida, elaborar um modelo matemático com o objetivo de predizer o ritmo de filtração glomerular (RFG) pós-operatório. Além disso, avaliar qual das equações mais usadas para estimativa do RFG por meio da creatinina sérica (Cockcroft-Gault, MDRD e CKD-EPI) tem melhor concordância com o 51Cr-EDTA. MÉTODOS: Este é um estudo prospectivo onde foram coletados dados demográficos, clínicos, laboratoriais e radiológicos pré-operatórios e seis meses após a nefrectomia, incluindo variáveis pouco ou nunca antes estudadas como a função renal diferencial na cintilografia com DMSA e o RFG mensurado por meio de estudo radioisotópico (51Cr-EDTA). Análises univariadas e multivariadas foram realizadas para identificar possíveis fatores de risco independentes para a piora da função renal. Essas variáveis foram então utilizadas na elaboração de um modelo matemático cujo objetivo foi estimar a função renal pós-operatória dos pacientes, utilizando para isso, apenas variáveis pré-operatórias. Por meio de estudos de correlação, os valores do RFG estimados pelas equações que utilizam a creatinina sérica foram comparados aos valores mensurados por meio do 51Cr-EDTA para avaliar qual tinha melhor concordância com o método padrão-ouro. RESULTADOS: De Abril de 2014 a Janeiro de 2018, 107 pacientes foram incluídos e completaram o protocolo de pesquisa. Doenças benignas foram responsáveis pela indicação da nefrectomia em 63,6% dos casos. Na análise univariada, diversas variáveis foram identificadas como possíveis fatores associados à evolução da função renal: idade, HAS, DM, DLP, DMSA e 51Cr-EDTA pré-operatórios. Entretanto, a análise multivariada demonstrou que a idade avançada (p=0,008), uma função relativa alta no DMSA do rim retirado (p < 0,001) e um valor de 51Cr-EDTA pré-operatório reduzido (p < 0,001) foram as variáveis que mantiveram significância e portanto foram consideradas fatores de risco independentes. A partir destas variáveis, elaborou-se um modelo matemático para estimativa do RFG pós-operatório (51Cr-EDTA pós-operatório = 37,9 - (0,29 x Idade) - (0,42 x DMSA RA) + (0,67 x 51Cr-EDTA pré-operatório). Por meio de análises de correlações, demonstrou-se que os valores do RFG obtidos por meio da equação CKD-EPI apresentavam melhor concordância com os valores mensurados com o 51Cr-EDTA. CONCLUSÕES: O presente protocolo de pesquisa demonstrou que a idade, o DMSA e o 51Cr-EDTA pré-operatórios estão significativamente associados à evolução da função renal após nefrectomias unilaterais e possibilitou a construção de um modelo para predizer o RFG pós-operatório. Também demonstrou que a equação CKD-EPI apresenta melhor concordância com o método considerado padrão-ouro para medida do RFG nesta população de pacientes / INTRODUCTION: The kidneys can be affected by pathologies that compromise their function in several degrees and unilateral nephrectomy is the best treatment option in many cases. However, there are still controversies about the renal function outcomes after nephrectomies and there are few retrospective studies that proposed models to estimate the postoperative glomerular filtration rate (GFR) after surgery. Moreover, they have methodological flaws and contradictory results. OBJECTIVES: To prospectively evaluate preoperative prognostic factors associated to renal function outcomes six months after unilateral nephrectomies with a gold-standard isotopic technique (51Cr-EDTA). To formulate a model for estimate the postoperative GFR. And to evaluate which equation for GFR estimation using serum creatinine has the best concordance to the 51Cr-EDTA. METHODS: Preoperative variables were prospectively collected and included: demographics, clinical, laboratorial and imaging studies. Univariate and multivariate analyses were done to identify the independent risk factors associated to renal function outcomes. These variables were used to create a model in order to predict the postoperative GFR. Correlation analyses were performed to evaluate which equation for GFR estimation using serum creatinine has the best concordance to the gold-standard isotope technique. RESULTS: One hundred and seven patients were enrolled and completed the study protocol from April 2014 to January 2018. Nephrectomy was performed for a benign disease in 63,2% of patients. After univariate and multivariate analyses, older age (p=0,008), higher split function of the affected kidney in DMSA scintigraphy (p < 0,001) and lower values of preoperative 51Cr-EDTA (p < 0,001) were identify as independent risk factors for postoperative GFR worsening. Using these variables, a mathematical model was elaborated to predict the postoperative GFR (postoperative 51Cr-EDTA = 37.9 - (0.29 x Age) - (0.42 x DMSA) + (0.67 x preoperative 51Cr-EDTA). Correlation analyses showed that GFR estimated by CKD-EPI equation has the best concordance to GFR measured by 51Cr-EDTA. CONCLUSIONS: The present study protocol demonstrated that age, DMSA and preoperative 51Cr-EDTA are significantly associated to postoperative renal function outcomes after unilateral nephrectomies and permitted the elaboration of a model to predict the postoperative GFR. Also demonstrated that CKD-EPI equation has the best concordance to the gold-standard technique for GFR measurement

Fundação de Amparo a Pesquisa do Estado de Minas Gerais / Os radionuclídeos utilizados na tomografia por emissão de posítrons (PET) são produzidos utilizando um acelerador cíclotron. Os nêutrons produzidos durante a operação do cíclotron contribuem para exposição direta ou indireta dos Indivíduos Ocupacionalmente Expostos (IOEs), devido ao aumento da radiação de fundo da casamata. Além disso, há um aumento nas emissões de gases radioativos provenientes da ativação dos elementos do ar dentro da casamata, que quando liberados constitui um problema para radioproteção dos indivíduos do público. Dos vários métodos utilizados para caracterizar o espectro neutrônico, o espectrômetro de multiesferas de Bonner (EB) é um dos sistemas espectrométricos mais utilizados. Neste trabalho foi utilizado o sistema EB com detectores termoluminescentes (TL), do tipo TLD600 e TLD700 como detector de nêutrons, para medir os espectros de energia de nêutrons em quatro pontos no interior da casamata do cíclotron GE PETtrace-8 do Centro de Desenvolvimento da Tecnologia Nuclear (CDTN). Foram realizadas medidas em quatro pontos em torno do cíclotron. Os espectros de nêutrons foram desdobrados utilizando os códigos BUMS, NSDUAZ e BUNKIUT e os resultados convertidos em equivalente de dose ambiente H*(10). Considerando o termo fonte de radiação fornecido pelo fabricante do cíclotron, pôde-se constatar a grande influência dos nêutrons de recuo nos espectros de energia encontrados em todos os pontos. Houve uma boa concordância nos espectros de nêutrons obtidos, utilizando os códigos BUNKIUT (com espectros iniciais retangular e Maxwelliano) e NSDUAZ. Os valores de taxa de fluência encontrados no presente trabalho foram da mesma magnitude dos valores reportados na literatura e são coerentes com os obtidos por cálculos téóricos utilizando o termo fonte de radiação disponibilizado pelo fabricante. Com relação aos valores de equivalente de dose ambiente, as taxas horárias por &#61549;A (microampère) variaram de aproximadamente 67 mSv/h a 936 mSv/h . Para uma corrente típica de 40 &#61549;A, esses valores são próximos de 2,7 Sv/h a 37 Sv/h, valores da mesma ordem dos reportados na literatura. A metodologia empregada para a caracterização do campo de radiação em torno do cíclotron do CDTN mostrou-se adequada e pode ser utilizada em mais pontos da casamata, de maneira a descrever melhor o espectro e, consequentemente, estimar o equivalente de dose ambiente. / The radionuclides used in positron emission tomography (PET) are generally produced using a cyclotron accelerator. The operation of the cyclotron produces an undesirable neutron radiation field. The knowledge of the neutron radiation field around not-self-shielded PET cyclotrons is an important issue for optimization of radiation protection of the workers and individuals of the public. For the workers, neutrons contribute not only for immediate radiation exposure as for long-term exposure due to activation of cyclotron components and the concrete in the bunker walls. For the individuals of the public the main concern is the dispersal of radioactive gases produced by activation of the air inside the cyclotron vault. The multisphere system, or Bonner sphere spectrometer (BSS), has been widely used to measure neutron spectrum. The substitution of the active detectors of the BSS system by thermoluminescent detectors (specifically TLD-600 and TLD-700 pairs) has become a reliable procedure in spectrometry of high intensity mixed radiation field. In this study we utilized the BSS system with TLD600 and TLD700 to measure the energy spectra of neutrons at four points inside the bunker of the cyclotron GE PETtrace-8 of the Development Centre of Nuclear Technology (CDTN). Four points inside the bunker of the cyclotron were studied. The neutron spectra were unfolded using codes BUMS, NSDUAZ e BUNKIUT and the results converted to ambient dose equivalent H*(10). Considering the source-term of radiation provided by the manufacturer of the cyclotron, we could see the great influence of room return effect in energy spectra at all points. The values of total fluence rates for all points have the same magnitudes of values reported in the literature and are consistent with those obtained by theoretical calculations using the source-term of radiation provided by the manufacturer of the cyclotron. The ambient equivalent dose rates for 1 &#61549;A ranged from about 67 mSv/h to 936 mSv/h. For a typical 40 &#61549;A typical current these values were 2.7 Sv/h and 37 Sv/h. These values are of the same order than the reported in the literature. The methodology utilized in this study to characterize the neutron radiation field around the CDTN cyclotron proved to be adequate and can be used in more points inside the bunker in order to better describe the spectrum and thereby estimate the ambient dose equivalent.

Radioisótopos de gálio são estudados e avaliados para aplicações médicas desde 1949. Nos últimos 50 anos 67Ga tem sido amplamente utilizado no diagnóstico de diversas patologias, incluindo lesões inflamatórias crônicas e agudas, bacterianas ou estéreis e diversos tipos de tumores. No Brasil 30% das clinicas que prestam serviços de Medicina Nuclear utilizam o Citrato de 67Ga com uma demanda de distribuição no IPEN-CNEN/SP de 37 GBq (1 Ci) por semana. O 67Ga apresenta meia-vida física de 3,26 dias (78 horas) e decai 100% por captura eletrônica para o 67Zn estável. Seu decaimento inclui a emissão de raios &gamma; com energias de 93,3 keV (37%), 184,6 keV (20,4%), 300,2 keV (16,6%) e 888 keV (26%). No IPEN o 67Ga era produzido a partir da reação 68Zn(p, 2n)67Ga. Após a irradiação, o alvo era totalmente dissolvido em HCl concentrado e a solução percolada em resina catiônica DOWEX 50W-X8, 200-400 mesh, condicionada em HCl 10 mol L-1. Zinco, níquel e cobre eram eluídos em HCl 10 mol L-1 e o 67Ga em HCl 3,5 mol L-1. O produto final era obtido na forma de citrato de 67Ga. Este trabalho apresenta um método inédito, rápido, direto e eficiente de separação química e obtenção de 67GaCl3 a partir da difusão térmica (aquecimento do alvo) aliada à extração em ácido acético concentrado. A purificação foi realizada por cromatografia de troca iônica. Realizou-se a eletrodeposição do zinco natural em placas de cobre niquelado como substrato e os depósitos de zinco obtidos foram aderentes ao substrato, levemente brilhantes e uniformes. Os alvos foram irradiados com prótons de 26 MeV e corrente integrada de 10 &mu;A.h. Após a irradiação, os alvos foram aquecidos a 300 °C por 2 horas e colocados em contato com ácido acético concentrado por 1 hora. O rendimento médio de extração de 67Ga obtido foi de (72±10)%. Esta solução foi evaporada e o resíduo foi retomado em NH4OH 0,5 mol L-1. O 67Ga foi purificado em resina catiônica Dowex 50WX8 em meio de NH4OH. A recuperação obtida foi de (98 ± 2) %, de 67Ga. O eluido foi evaporado e retomado em HCl 0,1 mol L-1. A pureza química foi verificada por ICP-OES encontrando-se (2 ± 1) &mu;g mL -1 de zinco. As concentrações de ferro, cobre e níquel foram inferiores ao limite de detecção do método e aos limites de utilização de 67Ga. A pureza radionuclídica foi verificada por espectroscopia-&gamma; utilizando um detector de germânio Hiper-Puro encontrando-se valor superior a (99,9%). Este método inédito permite a obtenção de 67Ga com alta pureza química, radioquímica e radionuclídica em condições de processamento menos agressivas e corrosivas que o método comumente utilizado. / Radioisotopes of gallium have been studied and evaluated for medical applications since 1949. Over the past 50 years 67Ga has been widely used in the diagnosis of various diseases, including acute and chronic inflammatory lesions, bacterial or sterile and several types of tumors. In Brazil 30% of clinics that provide services for Nuclear Medicine use 67Ga citrate and the demand for 67Ga at IPEN-CNEN/SP is 37 GBq (1 Ci)/week. The 67Ga presents physical half-life of 3.26 days (78 hours) and decays 100% by electron capture to stable 67Zn. Its decay includes the emission of &gamma; rays with energies of 93.3 keV (37%), 184.6 keV (20.4%), 300.2 keV (16.6%) and 888 keV (26%). In the past 67Ga was produced by the reaction 68Zn (p, 2n) 67Ga at IPEN-CNEN/SP. After irradiation, the target was dissolved in concentrated HCl and the solution percolated through a cationic resin DOWEX 50W-X8, 200-400 mesh, conditioned with 10 mol L -1 HCl. Zinc, nickel and copper were eluted in 10 mol L-1 HCl and 67Ga 3.5 mol L-1 HCl. The final product was obtained as 67Ga citrate. This work presents a new, fast, direct and efficient method for the chemical separation of 67Ga by thermal diffusion (heating of the target) combined with concentrated acetic acid extraction. Purification was performed by ion exchange chromatography. Natural zinc electrodeposition was performed on nickel/copper plates as substrate and the zinc deposits were adherent to the substrate, slightly shiny and uniform. The targets were irradiated with 26 MeV protons and integrated current of 10 &mu;A.h. After irradiation, the targets were heated at 300 °C for 2 hours and placed in contact with concentrated acetic acid for 1 hour. The average yield of extraction of 67Ga was (72 ± 10)%. This solution was evaporated and the residue was taken up in 0.5 mol L-1 NH4OH. The 67Ga was purified on cationic resin Dowex 50WX8 in NH4OH medium. The 67Ga recovery was (98 ± 2)%. This solution was evaporated and taken up in 0.1 mol L-1 HCl. The chemical purity was evaluated by ICP-OES that resulted in (2 ± 1) &mu;g mL-1 of zinc. The concentration of iron, copper and nickel was lower than the detection limits and also than the utilization limits for 67Ga. The radionuclidic purity was greater than (99.9%). This method showed to be suitable to obtain high purity 67Ga in less aggressive chemical conditions than before.

Radioisótopos de gálio são estudados e avaliados para aplicações médicas desde 1949. Nos últimos 50 anos 67Ga tem sido amplamente utilizado no diagnóstico de diversas patologias, incluindo lesões inflamatórias crônicas e agudas, bacterianas ou estéreis e diversos tipos de tumores. No Brasil 30% das clinicas que prestam serviços de Medicina Nuclear utilizam o Citrato de 67Ga com uma demanda de distribuição no IPEN-CNEN/SP de 37 GBq (1 Ci) por semana. O 67Ga apresenta meia-vida física de 3,26 dias (78 horas) e decai 100% por captura eletrônica para o 67Zn estável. Seu decaimento inclui a emissão de raios &gamma; com energias de 93,3 keV (37%), 184,6 keV (20,4%), 300,2 keV (16,6%) e 888 keV (26%). No IPEN o 67Ga era produzido a partir da reação 68Zn(p, 2n)67Ga. Após a irradiação, o alvo era totalmente dissolvido em HCl concentrado e a solução percolada em resina catiônica DOWEX 50W-X8, 200-400 mesh, condicionada em HCl 10 mol L-1. Zinco, níquel e cobre eram eluídos em HCl 10 mol L-1 e o 67Ga em HCl 3,5 mol L-1. O produto final era obtido na forma de citrato de 67Ga. Este trabalho apresenta um método inédito, rápido, direto e eficiente de separação química e obtenção de 67GaCl3 a partir da difusão térmica (aquecimento do alvo) aliada à extração em ácido acético concentrado. A purificação foi realizada por cromatografia de troca iônica. Realizou-se a eletrodeposição do zinco natural em placas de cobre niquelado como substrato e os depósitos de zinco obtidos foram aderentes ao substrato, levemente brilhantes e uniformes. Os alvos foram irradiados com prótons de 26 MeV e corrente integrada de 10 &mu;A.h. Após a irradiação, os alvos foram aquecidos a 300 °C por 2 horas e colocados em contato com ácido acético concentrado por 1 hora. O rendimento médio de extração de 67Ga obtido foi de (72±10)%. Esta solução foi evaporada e o resíduo foi retomado em NH4OH 0,5 mol L-1. O 67Ga foi purificado em resina catiônica Dowex 50WX8 em meio de NH4OH. A recuperação obtida foi de (98 ± 2) %, de 67Ga. O eluido foi evaporado e retomado em HCl 0,1 mol L-1. A pureza química foi verificada por ICP-OES encontrando-se (2 ± 1) &mu;g mL -1 de zinco. As concentrações de ferro, cobre e níquel foram inferiores ao limite de detecção do método e aos limites de utilização de 67Ga. A pureza radionuclídica foi verificada por espectroscopia-&gamma; utilizando um detector de germânio Hiper-Puro encontrando-se valor superior a (99,9%). Este método inédito permite a obtenção de 67Ga com alta pureza química, radioquímica e radionuclídica em condições de processamento menos agressivas e corrosivas que o método comumente utilizado. / Radioisotopes of gallium have been studied and evaluated for medical applications since 1949. Over the past 50 years 67Ga has been widely used in the diagnosis of various diseases, including acute and chronic inflammatory lesions, bacterial or sterile and several types of tumors. In Brazil 30% of clinics that provide services for Nuclear Medicine use 67Ga citrate and the demand for 67Ga at IPEN-CNEN/SP is 37 GBq (1 Ci)/week. The 67Ga presents physical half-life of 3.26 days (78 hours) and decays 100% by electron capture to stable 67Zn. Its decay includes the emission of &gamma; rays with energies of 93.3 keV (37%), 184.6 keV (20.4%), 300.2 keV (16.6%) and 888 keV (26%). In the past 67Ga was produced by the reaction 68Zn (p, 2n) 67Ga at IPEN-CNEN/SP. After irradiation, the target was dissolved in concentrated HCl and the solution percolated through a cationic resin DOWEX 50W-X8, 200-400 mesh, conditioned with 10 mol L -1 HCl. Zinc, nickel and copper were eluted in 10 mol L-1 HCl and 67Ga 3.5 mol L-1 HCl. The final product was obtained as 67Ga citrate. This work presents a new, fast, direct and efficient method for the chemical separation of 67Ga by thermal diffusion (heating of the target) combined with concentrated acetic acid extraction. Purification was performed by ion exchange chromatography. Natural zinc electrodeposition was performed on nickel/copper plates as substrate and the zinc deposits were adherent to the substrate, slightly shiny and uniform. The targets were irradiated with 26 MeV protons and integrated current of 10 &mu;A.h. After irradiation, the targets were heated at 300 °C for 2 hours and placed in contact with concentrated acetic acid for 1 hour. The average yield of extraction of 67Ga was (72 ± 10)%. This solution was evaporated and the residue was taken up in 0.5 mol L-1 NH4OH. The 67Ga was purified on cationic resin Dowex 50WX8 in NH4OH medium. The 67Ga recovery was (98 ± 2)%. This solution was evaporated and taken up in 0.1 mol L-1 HCl. The chemical purity was evaluated by ICP-OES that resulted in (2 ± 1) &mu;g mL-1 of zinc. The concentration of iron, copper and nickel was lower than the detection limits and also than the utilization limits for 67Ga. The radionuclidic purity was greater than (99.9%). This method showed to be suitable to obtain high purity 67Ga in less aggressive chemical conditions than before.

Στην παρούσα Μεταπτυχιακή Διπλωματική Εργασία μελετάται το πρόβλημα των θωρακίσεων ακτινοπροστασίας σε τρεις χαρακτηριστικούς χώρους ενός εργαστηρίου Πυρηνικής Ιατρικής, σύμφωνα με τους Κανονισμούς Ακτινοπροστασίας. Οι υπολογισμοί έγιναν με την εφαρμογή λογιστικών φύλλων Microsoft Excel. Ο σκοπός των κάθε είδους θωρακίσεων ιοντιζουσών ακτινοβολιών είναι η μείωση της δόσης της ακτινοβολίας στην οποία εκτίθενται και απορροφούν οι εργαζόμενοι, οι ασθενείς/εξεταζόμενοι αλλά και οι απλοί επισκέπτες σε χώρους ακτινοβόλησης ή γειτονικών, μέσα στα επιτρεπτά όρια. Στην μελέτη αυτή υπολογίζεται το πάχος θωράκισης που χρειάζεται να τοποθετηθεί σε έναν χώρο έτσι ώστε να μην γίνεται υπέρβαση των Περιοριστικών Επιπέδων Δόσεων (Π.Ε.Δ.). Τα πιο κοινά υλικά θωράκισης είναι ο μόλυβδος, το μπετό/σκυρόδεμα και ο σίδηρος. Ο πρώτος χώρος περιέχει ένα ραδιοϊσότοπο, μέσα σε κρύπτη, ενώ υπολογίζεται και ο ρυθμός δόσης σε ένα άτομο, το οποίο μπορεί να βρίσκεται είτε μέσα στον ίδιο χώρο είτε σε παρακείμενο. Ο δεύτερος χώρος περιέχει δύο ραδιοϊσότοπα, και υπολογίζεται η συνολική θωράκιση που απαιτείται. Ο τρίτος χώρος είναι ο χώρος αναμονής ενός πραγματικού εργαστηρίου Πυρηνικής Ιατρικής, μέσα στον οποίο μπορούν να υπάρχουν ταυτόχρονα μέχρι επτά ασθενείς, σε κάθε έναν από τους οποίους έχει χορηγηθεί το απαραίτητο ραδιοφάρμακο για την δική του εξέταση. Επιπλέον, λαμβάνεται υπ’ όψιν η ενδοαπορρόφηση σε κάθε ασθενή, θεωρώντας ότι αποτελείται από έναν κύλινδρο (το σώμα) και μια σφαίρα (το κεφάλι). Ο χώρος αναμονής περιβάλλεται από τον διάδρομο, την αίθουσα αιμοληψιών, το θερμό εργαστήριο, το δωμάτιο εφαρμογής και ο χώρος της γ-κάμερας. Η εφαρμογή λογιστικών φύλλων Microsoft Excel επελέγη για την υλοποίηση των υπολογισμών, ούτως ώστε οι εξισώσεις και οι υπολογισμοί να είναι ανοικτοί και εύκολα επαληθεύσιμοι από τον χρήστη. Επιπλέον, το πακέτο Microsoft Excel καθώς και η λειτουργία του είναι ευρέως διαδεδομένα. Ο χρήστης έχει πλήρη έλεγχο σε κάθε παράμετρο του κάθε χώρου, όπως π.χ. διαστάσεις του χώρου, ραδιοϊσότοπο και ενεργότητα, εξέταση, μέγεθος ασθενούς, κατηγορία παρακείμενων χώρων, υλικό θωράκισης κτλ. Η εφαρμογή διαβάζει αυτόματα ό,τι πληροφορίες απαιτούνται (π.χ. Περιοριστικά Επίπεδα Δόσης, Half-Value Layer κτλ.) από τον πίνακα δεδομένων, ο οποίος και αυτός μπορεί να ενημερωθεί ή εμπλουτιστεί από τον χρήστη, και υπολογίζει τις ζητούμενες θωρακίσεις. Τέλος, η εφαρμογή έχει σχεδιαστεί έτσι ώστε να είναι ευέλικτη και να μπορεί εύκολα να χρησιμοποιηθεί για άλλους χώρους και εργαστήρια, είτε ως έχει είτε με μικρές τροποποιήσεις. / This Master Thesis studies the problem of radiation protection shielding in three typical areas of a Nuclear Medicine Laboratory, in accordance with Radiation Protection Regulations. The actual calculations are performed using the spreadsheet software package Microsoft Excel. The purpose of any type of ionizing radiation shielding is to reduce to within the allowable limits the dose of radiation that workers, patients and ordinary visitors are exposed to and absorb either in radiation areas or in adjacent ones. In this study the thickness of shielding which needs to be placed in an area so as not to exceed the Dose Constraints is calculated. The most common shielding materials are lead, concrete and iron. The first area contains a radioisotope within a crypt, and the dose rate is calculated to an individual, who may be located either within the same room or in an adjacent one. The second area contains two radioisotopes, and the required total shielding is calculated. The third area is the waiting room of an actual Nuclear Medicine laboratory, in which up to seven patients, each of whom has been administered the necessary radiopharmaceutical for his examination, can exist simultaneously. The internal absorption of each patient is taken into account, modeling the patient as consisting of a cylinder (the body) and a sphere (the head). The waiting room is surrounded by a corridor, the blood sampling room, the hot lab, the radionuclide administration area, and the gamma camera area. The spreadsheet application Microsoft Excel was chosen for the implementation of the calculations, so that the equations and calculations be open and easily verifiable by the user. In addition, Microsoft Excel and its use are widespread. The user has full control over every aspect of each area, e.g. dimensions of space, radioisotope and activity, examination, patient size, category of adjacent area, shielding material etc. The application automatically reads the required information (e.g. Dose Constraints, Half-Value Layer etc.) from the data table, which may also be updated and enriched by the user, and calculates the required shielding. Finally, the application is designed to be flexible and can easily be used for other areas and laboratories, either as it is or with minor modifications.

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