Your browser doesn't support javascript.
loading
Mostrar: 20 | 50 | 100
Resultados 1 - 2 de 2
Filtrar
Mais filtros

Bases de dados
País/Região como assunto
Ano de publicação
Tipo de documento
País de afiliação
Intervalo de ano de publicação
1.
Radiat Prot Dosimetry ; 180(1-4): 42-45, 2018 Aug 01.
Artigo em Inglês | MEDLINE | ID: mdl-29518232

RESUMO

This study describes the use of a neutron irradiator system based on a plutonium-beryllium neutron source for MnSO4 solution activation for use to determine the MSB system efficiency. Computational simulations using Monte Carlo code were performed to establish the main characteristics of the irradiator system. Among the simulated geometries and volumes, semi-cylindrical shape with 84.5 cm3 of MnSO4 solution yielded the best option to be built. Activity measurements were performed with a high-pure germanium detector to validate the new irradiation system. Results showed an average efficiency and uncertainty of the experimental standard deviation of the mean: 5.742 × 10-4 ± 0.036 × 10-4 (coverage factor k = 1), for MSB system. Efficiency value obtained shows good correlation to other published methods (i.e. a relative difference of 0.07%). This alternative metrological method demonstrated the utility of neutron sources for the irradiation of solutions in metrology laboratories providing a cost-efficient alternative to nuclear reactors or particle accelerators.


Assuntos
Berílio/análise , Compostos de Manganês/química , Nêutrons , Plutônio/análise , Monitoramento de Radiação/instrumentação , Sulfatos/química , Calibragem , Método de Monte Carlo , Doses de Radiação
2.
Radiat Prot Dosimetry ; 161(1-4): 185-9, 2014 Oct.
Artigo em Inglês | MEDLINE | ID: mdl-24625545

RESUMO

The Brazilian Metrology Laboratory of Ionizing Radiations (LNMRI) standard thermal neutron flux facility was designed to provide uniform neutron fluence for calibration of small neutron detectors and individual dosemeters. This fluence is obtained by neutron moderation from four (241)Am-Be sources, each with 596 GBq, in a facility built with blocks of graphite/paraffin compound and high-purity carbon graphite. This study was carried out in two steps. In the first step, simulations using the MCNPX code on different geometric arrangements of moderator materials and neutron sources were performed. The quality of the resulting neutron fluence in terms of spectrum, cadmium ratio and gamma-neutron ratio was evaluated. In the second step, the system was assembled based on the results obtained on the simulations, and new measurements are being made. These measurements will validate the system, and other intercomparisons will ensure traceability to the International System of Units.


Assuntos
Proteção Radiológica/métodos , Radiometria/instrumentação , Radiometria/métodos , Amerício , Berílio , Brasil , Calibragem , Carbono/química , Simulação por Computador , Desenho de Equipamento , Grafite/química , Método de Monte Carlo , Nêutrons , Doses de Radiação , Radiação Ionizante
SELEÇÃO DE REFERÊNCIAS
DETALHE DA PESQUISA