Absorbed dose comparison: positron emitters 11 C, 13 N, and 15 O versus gamma-ray emitters

1979 ◽  
Vol 6 (3) ◽  
pp. 221-223 ◽  
Author(s):  
Paul A. Feller ◽  
Vincent J. Sodd ◽  
Hiroshi Nishiyama
Keyword(s):  
Gamma Ray ◽  
Absorbed Dose ◽  
Dose Comparison ◽  
Positron Emitters ◽  
Gamma Ray Emitters

scholarly journals Monte carlo simulation of different positron emitting radionuclides incorporated in a soft tissue volume

2017 ◽  
Vol 8 (1) ◽  
Author(s):  
Segundo Agustín Martínez Ovalle
Keyword(s):  
Monte Carlo Simulation ◽  
Monte Carlo ◽  
Soft Tissue ◽  
Gamma Ray ◽  
Absorbed Dose ◽  
Tissue Volume ◽  
Monte Carlo Calculations ◽  
Soft Tissue Volume ◽  
Maximum Range ◽  
Positron Emitters

Monte Carlo calculations were carried out where compounds with positron-emitters radionuclides, like FDG (18F), Acetate (11C), and Ammonium (13N), were incorporated into a soft tissue volume, in the aim to estimate the type of particles produced their energies, their mean free paths, and the absorbed dose at different distances with respect to the center of the volume. The volume was modeled with a radius larger than the maximum range of positrons in order to produce 0.511 keV annihilation gamma-ray photons. With the obtained results the absorbed dose, in various organs and tissues able to metabolize different radiopharmaceutical drugs, can be estimated. The code used was GEANT4.

Development of SiO2 based doped with LiF, Cr2O3, CoO4 and B2O3 glasses for gamma and fast neutron shielding

2020 ◽  
Vol 0 (0) ◽  
Author(s):  
Bünyamin Aygün ◽  
Erdem Şakar ◽  
Abdulhalik Karabulut ◽  
Bünyamin Alım ◽  
Mohammed I. Sayyed ◽  
...  
Keyword(s):  
Monte Carlo Simulation ◽  
Monte Carlo ◽  
Fast Neutron ◽  
Cross Sections ◽  
Gamma Ray ◽  
Absorbed Dose ◽  
Neutron Shielding ◽  
Shielding Properties ◽  
Effective Neutron ◽  
Simulation Codes

AbstractIn this study, the fast neutron and gamma-ray absorption capacities of the new glasses have been investigated, which are obtained by doping CoO,CdWO4,Bi2O3, Cr2O3, ZnO, LiF,B2O3 and PbO compounds to SiO2 based glasses. GEANT4 and FLUKA Monte Carlo simulation codes have been used in the planning of the samples. The glasses were produced using a well-known melt-quenching technique. The effective neutron removal cross-sections, mean free paths, half-value layer, and transmission numbers of the fabricated glasses have been calculated through both GEANT4 and FLUKA Monte Carlo simulation codes. Experimental neutron absorbed dose measurements have been carried out. It was found that GS4 glass has the best neutron protection capacity among the produced glasses. In addition to neutron shielding properties, the gamma-ray attenuation capacities, were calculated using newly developed Phy-X/PSD software. The gamma-ray shielding properties of GS1 and GS2 are found to be equivalent to Pb-based glass.

Natural radioactivity in some building materials and assessment of the associated radiation hazards

2018 ◽  
Vol 106 (1) ◽  
pp. 79-86
Author(s):  
Amira Kasumović ◽  
Ema Hankić ◽  
Amela Kasić ◽  
Feriz Adrović
Keyword(s):  
Building Materials ◽  
Gamma Ray ◽  
Absorbed Dose ◽  
Absorbed Dose Rate ◽  
Radium Equivalent Activity ◽  
Radiation Hazards ◽  
Eu Member States ◽  
Gamma Ray Spectrometer ◽  
Radioactive Nuclides ◽  
Specific Activities

AbstractThe results of the specific activities of232Th,226Ra and40K measured in samples of commonly used building materials in Bosnia and Herzegovina are presented. Measurements were performed by gamma-ray spectrometer with coaxial HPGe detector. The surface radon exhalation and mass exhalation rates for selected building materials were also measured. The determined values of specific activities were in range from 3.16±0.81 Bq kg−1to 64.79±6.16 Bq kg−1for232Th, from 2.46±0.95 Bq kg−1to 53.89 ±3.67 Bq kg−1for226Ra and from 28.44±7.28 Bq kg−1to 557.30±93.38 Bq kg−1for40K. The radium equivalent activity, the activity concentration index, the external and internal hazard indices as well as the absorbed dose rate in indoor air and the corresponding annual effective dose, due to gamma-ray emission from the radioactive nuclides in the building material, were evaluated in order to assess the radiation hazards for people. The measured specific activities of the natural radioactive nuclides in all investigated building materials were compared with the published results for building materials from other European countries. It can be noted that the results from this study are similar to the data for building materials from neighbouring countries and for building materials used in the EU Member States. The radiological hazard parameters of the building materials were all within the recommended limits for safety use.

Development of two dosimetry systems based on basic violet dye for possible use as radiation dosimeters

2021 ◽  
Vol ahead-of-print (ahead-of-print) ◽  
Author(s):  
Nehad Magdy ◽  
Sameh Gafar
Keyword(s):  
Gamma Ray ◽  
Raw Materials ◽  
Absorbed Dose ◽  
Chemical Yield ◽  
Radiation Sensitivity ◽  
Vital Role ◽  
Radiation Chemical ◽  
Content Type ◽  
Before And After

Purpose The purpose of this research paper is to study a comparison between two dosimetry systems, both of them based on basic violet dye (BV). Design/methodology/approach The first system depends on (BV) (incorporating polyvinyl alcohol) as a thin-film dosimeter. The second system also relies on (BV) as a solution dosimeter, which is more sensitive to gamma rays. The two prepared film/solutions have a considerable signal that decreases upon irradiation and the strength of the signal decreases with increasing radiation dose. Findings The gamma ray absorbed dose for these dosimeters was found to be up to 35 kGy for films and 1 kGy for the liquid phase. All dosimetric characteristics as radiation chemical yield, additive substance, dose-response function, radiation sensitivity, also before and after-irradiation stability under various conditions were considered. Practical implications It is expected the vital role of gamma radiation on this dye in its two forms or two media. This reveals their wide applications in the field of gamma irradiation processing. Originality/value These two dosimetry systems which depend upon the same dye are safe to handle, inexpensive, available raw materials and can be applied in various dosimetry applications as mentioned above.

ESTIMATION OF RADIOLOGICAL HAZARDS DUE TO NATURAL RADIONUCLIDES FROM THE ROSTERMAN GOLD MINE TAILINGS, LURAMBI, KAKAMEGA, KENYA

2020 ◽  
Vol 190 (3) ◽  
pp. 324-330
Author(s):  
C K Wanyama ◽  
F W Masinde ◽  
J W Makokha ◽  
S M Matsitsi
Keyword(s):  
Gamma Ray ◽  
Natural Radionuclides ◽  
Absorbed Dose ◽  
Spectrometric Analysis ◽  
Gold Mine ◽  
Absorbed Dose Rate ◽  
Radiological Hazards ◽  
Naturally Occurring Radionuclides ◽  
Gold Mine Tailings ◽  
Radiological Health

Abstract Radiological hazards associated with naturally occurring radionuclides in materials from Rosterman gold mine were assessed by analysis of 30 samples. The gamma-ray spectrometric analysis of tailing samples reported an average activity concentration of 263 ± 13, 123 ± 6 and 84 ± 4 Bq kg−1 for 40K, 232Th and 226Ra, respectively. The average absorbed dose rate was 124 ± 6 nGy h−1, while the annual effective dose of 0.4 ± 0.02 mSv y−1 for indoor and 0.3 ± 0.01 mSv y−1 for outdoor were reported. The mean and range of radiological parameters (external and radium equivalent) calculated from the tailing samples were within the permissible limits and hence mining of gold at Rosterman has no significant radiological health implication on the miners and the surrounding population.

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