Gamma-ray Dosimetry of Internal Emitters. Monte Carlo Calculations of Absorbed Dose from Point Sources

1964 ◽  
Vol 37 (433) ◽  
pp. 45-52 ◽  
Author(s):  
W. H. Ellett ◽  
A. B. Callahan ◽  
G. L. Brownell
Keyword(s):  
Monte Carlo ◽  
Gamma Ray ◽  
Absorbed Dose ◽  
Point Sources ◽  
Monte Carlo Calculations

Gamma-ray Dosimetry of Internal Emitters II: Monte Carlo Calculations of Absorbed Dose from Uniform Sources

1965 ◽  
Vol 38 (451) ◽  
pp. 541-544 ◽  
Author(s):  
William H. Ellett ◽  
Arthur B. Callahan ◽  
Gordon L. Brownell
Keyword(s):  
Monte Carlo ◽  
Gamma Ray ◽  
Absorbed Dose ◽  
Monte Carlo Calculations

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.

scholarly journals Characterization and Monte Carlo simulations for a CLYC detector

2018 ◽  
Vol 48 ◽  
pp. 1860115
Author(s):  
Alessandro Borella ◽  
Eric Boogers ◽  
Riccardo Rossa ◽  
Peter Schillebeeckx
Keyword(s):  
Monte Carlo ◽  
Gamma Ray ◽  
Detection System ◽  
Nuclear Material ◽  
Point Sources ◽  
Pulse Shape Discrimination ◽  
Monte Carlo Code ◽  
Nuclear Security ◽  
Collaborative Agreement ◽  
Personal Dosimetry

The CLYC (Cs[Formula: see text]LiYCl[Formula: see text]:Ce) detector is a scintillator detector sensitive to both neutron and gamma radiation and capable of separating the two types of radiation by pulse-shape discrimination. This feature is interesting as pertains to the development of non-destructive assays for the safeguard of nuclear material, nuclear security, and fast-neutron personal dosimetry. A [Formula: see text] CLYC detector highly enriched with [Formula: see text]Li was purchased and tested with analog and digital electronics. In this work, we report on the characterization of the detector in terms of linearity, energy resolution, and full-energy efficiency for gamma rays. This characterization was achieved by measurements with calibrated gamma-ray point-sources with an analog measuring chain, in a well-defined, reproducible geometry. The experimental data were also used to validate a model of the detection system that was developed with the Monte Carlo code MCNP-CP. This work is part of a collaborative agreement between SCK•CEN and JRC-Geel.

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