Este trabalho descreve a modelagem computacional de dispersão atmosférica decorrente de acidente radiológico hipotético em reator modular de pequeno porte (SMR), cuja potência é de 16 MWe (50 MWt). Utilizou-se o software SCALE para modelar o núcleo com três regiões de enriquecimento do combustível, a 4%, 5% e 20%, e obter atividades dos radionuclídeos oriundos de reações nucleares durante o burnup, após 2 anos de operação. Foi escolhida uma localidade de interior para instalação do SMR, onde informações sobre condições meteorológicas foram coletadas para identificação da classe de estabilidade atmosférica predominante. Dentre os radionuclídeos do inventário, considerou-se a contribuição do Cs-137 para simulação, usando-se o código HotSpot, da concentração e das doses totais efetivas (TEDE) recebidas, ambas em função da distância do evento. Os resultados sugerem que a TEDE máxima calculada foi de 3,6 Sv, a 34 m do reator, diminuindo com o tempo e distância, e seguindo o modelo Gaussiano de dispersão, e que a pluma de contaminação é dependente dos critérios de Pasquill-Gifford e da atividade do Cs-137. Para doses entre 1 mSv e 10 mSv e entre 10 mSv e 50 mSv, sugere-se a abrigagem da população nas construções existentes na localidade, e para valores acima de 50 mSv, a abrigagem nessas condições ou a evacuação do pessoal das proximidades do reator em movimento contrário ao de propagação da pluma. A relevância dessa investigação mostra a importância do planejamento de respostas em situação de emergência e a influência das condições meteorológicas, considerando-se os dados assumidos na
In the present work, the transmission factors of γ-rays are determined in bi-layered shields composed of lead and steel, through a methodology composed of three distinct parts. The buildup calculation was performed using the methodology published by Broder in 1962 [1]. A computational simulation was used through a spherical model, a total of three concentric spheres were simulated, with the source in the center of the spheres. The first sphere represents the lead shield and its radius is represented by the thickness of this material. The second sphere represents the steel shield and its radius is the sum of the thicknesses of the shielding. The third sphere is the vacuum that will determine the number of photons that will pass. To verify if the analytical methodology can be used to calculate the transmission factor of the proposed shield, laboratory experiments were performed with the BGO (Bismuth Germanate) detector. Measurements were only made with the thickness of steel, and with 15 different thicknesses of lead, ranging from 0.11 cm to 2.01 cm, while keeping the steel thickness. Three different thicknesses of steel were used: 0.65 cm, 0.85 cm and 1.40 cm. The work is relevant in the field of radiological and nuclear defense, considering the application of this shield in military vehicles, and the efficiency of the proposed analytical methodology was demonstrated.
Particle accelerator technology has a deep impact on society. Its applications are well established mainly in the treatment of cancer and other diseases. This work aims to develop an experimental apparatus with 3He detectors for 16 MeV photoneutron measurements. The apparatus allows us to obtain multienergetic neutrons with the use of a 22 cm diameter spherical attenuator associated with different shield thicknesses. The microscopic processes of the fast and thermal neutrons in the detector were described by the two energy-group diffusion equation. The Detector Response Dose Rate results show a directly proportional relationship between these two variables with a degree of reliability attested by the linear correlation coefficient .Particle accelerator technology has a deep impact on society. Its applications are well established mainly in the treatment of cancer and other diseases. This work aims to develop an experimental apparatus with 3He detectors for 16 MeV photoneutron measurements. The apparatus allows us to obtain multienergetic neutrons with the use of a 22 cm diameter spherical attenuator associated with different shield thicknesses. The microscopic processes of the fast and thermal neutrons in the detector were described by the two energy-group diffusion equation. The Detector Response Dose Rate results show a directly proportional relationship between these two variables with a degree of reliability attested by the linear correlation coefficient .
Purpose: Characterize the value of dose rate received by the nuclear medicine professionals during the examination procedure and handling of radiopharmaceuticals, permiting the realization of more specific pratices of radioprotection in nuclear medicine. Method and Materials: For this work was monitoring two professional in two separate nuclear medicine services, totaling four profissionals monitoring during two moths. Their routine was not modified. The measuring of the rate dose was made with an ionization chamber model Babyline 81 at a distance of 1.0 m from the source. The measurements were made during the eluation of radioisotope, preparation the radiopharmaceuticals, in the administration and during the exam realization. The value of the rate dose was associated the procedure realized, for this way to caracterize the level of exposure that the professional it was sudmitted during your daily work. Results: It was observed for the test of renal scintigraphy with DTPA an average of 2.55 μGy/h and an average value of 1.20 μGy/h to DMSA. For examinations of bone scintigraphy with MDP was observed an average of 2.63 μGy/h for the protocol of time and 3.09 μGy/h for the protocol of counting, the difference in protocol is due to different interpretation medical. The bone flow scintigraphy the average rate dose was 5.17 μGy/h. The values of the background radiation measured over the handling of radiopharmaceuticals shows a maximum increase of up to six times the background radiation before manipulation and this value influence the measurements performed by measuring activity. Conclusion: The evaluation of the values of dose rate may help in taking action directed to optimizing the radioprotection and minimize the sum of errors during a correct administered radiopharmaceutical to the patient, minimizing the dose in the pacient and in the professional.
O ser humano está sujeito diariamente a várias fontes de radiação natural, sendo a principal delas o radônio, pertencente à cadeia de decaimento radioativo do urânio e do tório [1]. No local de estudo, estão armazenadas aproximadamente 3,3 toneladas de urânio com concentração natural. Fato este, que gerou a preocupação quanto aos níveis de dose ocupacional dos IOE's da instalação, pois sua dose efetiva terá o acréscimo devido a dose equivalente proveniente da exposição ao radônio. As medições ocorreram: no depósito de urânio; numa sala contigua e no ponto de controle. De acordo com a Norma CNEN -NN 3.01 e considerando as concentrações de radônio existentes nos três ambientes concluiu-se que: a sala de depósito de urânio é considerada uma área controlada; o ponto de controle uma área livre e a sala contígua, uma área supervisionada [2].
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