Development of a Monte Carlo Code for the Calculation of Gamma Ray Transport in the Natural Environment

1985 ◽  
Vol 12 (1) ◽  
pp. 21-28 ◽  
Author(s):  
K. Saito ◽  
S. Moriuchi
Keyword(s):  
Monte Carlo ◽  
Natural Environment ◽  
Gamma Ray ◽  
Monte Carlo Code

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.

Neutron Radiography System Collimator Design via Monte Carlo Calculation

2010 ◽  
Author(s):  
Hamid Jafari ◽  
Majid Shahriari
Keyword(s):  
Monte Carlo ◽  
Neutron Flux ◽  
Neutron Source ◽  
Gamma Ray ◽  
Neutron Radiography ◽  
Image Plane ◽  
Monte Carlo Code ◽  
Radiography System ◽  
Neutron Filter ◽  
Non Destructive

Neutron radiography uses the unique interaction probabilities of neutrons to create images of materials. This imaging technique is non-destructive. MCNP Monte Carlo Code has been used to design an optimized neutron radiography system that utilizes 241Am-Be neutron source. Many different arrangements have been simulated to obtain a neutron flux with higher amplitude and more uniform distribution in the collimator outlet, next to image plane. In the final arrangement the specifications of neutron filter, Gamma-ray shield and beam collimator has been determined. Simulations has been Carried out for a 5Ci 241Am-Be neutron source. In this case 43.8 n/cm2s thermal neutron flux has been achieved at a distance of 35cm from neutron source.

Validation of gamma-ray true coincidence summing effects modeled by the Monte Carlo code MCNP-CP

2008 ◽  
Vol 278 (2) ◽  
pp. 359-363 ◽  
Author(s):  
H. Zhu ◽  
R. Venkataraman ◽  
N. Menaa ◽  
W. Mueller ◽  
S. Croft ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Gamma Ray ◽  
Monte Carlo Code ◽  
True Coincidence ◽  
True Coincidence Summing ◽  
Coincidence Summing

scholarly journals Gantry orientation effect on the neutron and capture gamma ray dose equivalent at the maze entrance door in radiation therapy

2012 ◽  
Vol 27 (1) ◽  
pp. 70-74 ◽  
Author(s):  
Hosein Ghiasi ◽  
Asghar Mesbahi
Keyword(s):  
Monte Carlo ◽  
Radiation Therapy ◽  
Gamma Ray ◽  
Mean Value ◽  
Gamma Dose ◽  
Monte Carlo Code ◽  
Dose Calculations ◽  
Entrance Door ◽  
Therapy Room ◽  
Gantry Angle

In the present study, the effect of gantry orientation on the photoneutron and capture gamma dose calculations for maze entrance door was evaluated. A typical radiation therapy room made of ordinary concrete was simulated using MCNPX Monte Carlo code. Gantry rotation was simulated at eight different angles around the isocenter. Both neutron and capture gamma dose vary considerably with gantry angle. The ratios of the maximum to the minimum values for neutron and capture gamma dose equivalents were 1.9 and 1.4, respectively. On the other hand, comparison of the Monte Carlo calculated mean value over all orientations with Monte Carlo calculated neutron and gamma dose showed that the Wu-McGinley method differed by 5% and 2%, respectively. However, for more conservative shielding calculations, factors of 1.6 and 1.3 should be applied to the calculated neutron and capture gamma doses at downward irradiation. Finally, it can be concluded that the gantry angle influences neutron and capture gamma dose at the maze entrance door and it should be taken into account in shielding considerations.

Comparison of mass attenuation coefficients of concretes using FLUKA, XCOM and experiment results

2018 ◽  
Vol 53 (2) ◽  
pp. 145-148 ◽  
Author(s):  
V.P. Singh ◽  
T. Korkut ◽  
N.M. Badiger
Keyword(s):  
Monte Carlo ◽  
Gamma Ray ◽  
High Energy ◽  
Steel Scrap ◽  
Linear Attenuation Coefficient ◽  
Monte Carlo Code ◽  
Attenuation Coefficients ◽  
Energy Gamma ◽  
Mass Attenuation Coefficients ◽  
Mass Attenuation

The mass attenuation coefficients of seven different types of normal and heavy concretes like ordinary, hematite-serpentine, ilmenite-limonite, basalt-magnetite, ilmenite, steel-scrap and steel-magnetite concretes has been simulated using FLUKA Monte Carlo code at high energies 1.5, 2, 3, 4, 5 and 6 MeV. The mass attenuation coefficients and linear attenuation coefficient of the concretes were found dependent upon the chemical composition, density and gamma ray energy. FLUKA Monte Carlo code results were found in good agreement with experimental and theoretical XCOM data. Our investigations for high energy gamma-ray interaction validate the FLUKA Monte Carlo code for use where experimental gamma-ray interaction results are not available.

scholarly journals HPGe Detector Energy Response Function Calculation Up to 400 keV Based on Monte Carlo Code

2010 ◽  
Vol 2 (3) ◽  
pp. 479 ◽  
Author(s):  
M. S. Rahman ◽  
G. Cho
Keyword(s):  
Monte Carlo ◽  
Response Function ◽  
Gamma Ray ◽  
Germanium Detector ◽  
Front Surface ◽  
Configuration Model ◽  
Monte Carlo Code ◽  
Response Functions ◽  
Energy Response ◽  
Measured Energy

A high purity germanium detector (HPGe) of crystal size of diameter 4.91 cm and length of 3.6 cm  was modeled in accordance with the Pop Top cryostat configuration (model no. GEM10P4). The energy response function was calculated in the air using Monte Carlo simulation with mono-energy g-ray photon up to 400 keV. The distance between the source and the front surface or end cap to detector was 20 cm and the source was assumed as an isotopic point source. The aluminum absorbing layers of thickness 0.127 cm was also taken into consideration in the simulation model. The input number of particles was 107 for each mono-energetic g-ray photon. The simulated energy response functions were verified with the measured energy response functions obtained using calibration sources in order to prove the accuracy of the modeling. The comparison between the measured energy response functions and the simulated energy response functions after normalization were also performed.  Keywords: HPGe; Gamma-ray spectrum; Monte Carlo. © 2010 JSR Publications. ISSN: 2070-0237 (Print); 2070-0245 (Online). All rights reserved. DOI: 10.3329/jsr.v2i3.4668                 J. Sci. Res. 2 (3), 479-483 (2010)  

Modeling of Cerenkov-Based Fiber-Optic Gamma-Ray Radiation Sensor Using Monte Carlo Simulation

2018 ◽  
Author(s):  
Hwa Jeong Han ◽  
Byung Gi Park ◽  
Beom Kyu Kim ◽  
Ji Hye Park ◽  
Won Ki Kim
Keyword(s):  
Monte Carlo ◽  
Optical Fiber ◽  
Gamma Rays ◽  
Gamma Ray ◽  
Fiber Optic ◽  
Monte Carlo Model ◽  
Signal Propagation ◽  
Optical Signal ◽  
Radiation Environment ◽  
Monte Carlo Code

In this study, a Monte Carlo model has been developed for a Cerenkov-based fiber-optic gamma-ray sensor (CFOGRS) using the GEANT4 simulation toolkit. The detection material for gamma rays in CFOGRS is the transparent silica core of the optical fiber, which is also used for optical signal propagation. The model implemented with the GEANT4 includes the transport process of gamma rays, as well as the physical processes of Compton scattering, the Cerenkov effect, and optical photon propagation within the optical fiber. The model also simulated the applicability of the CFOGRS in a radiation environment by using the Monte Carlo code of GEANT4.

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