Results 181 to 190 of about 6,588 (195)

Comparison between codes MCNPX and Gate/Geant4 in volume fraction studies

Applied Radiation and Isotopes, 2020
Knowing the volume fraction in a multiphase flow is of fundamental importance in predicting the performance of many systems and processes, it has been possible to model an experimental apparatus for volume fraction studies using Monte Carlo codes. Artificial neural networks have been applied for the recognition of the pulse height distributions in ...
Renato Raoni Werneck Affonso, Caroline Mattos Barbosa, Roos S.F. Dam, William L. Salgado, Ademir X. da Silva, César M. Salgado   +5 more
openaire   +2 more sources

Proton Therapy of eye using MCNPX code

2018
Introduction: Proton radiotherapy is the one of advanced teletherapy methods. The protons deposit their maximum energy in a position called Bragg peak. Therefore, for treatment of cancer, the tumor should be placed at the Bragg peak or SOBP. The scattered photons and neutrons is a challenge in proton radiotherapy.
Safaiean, S., Salehi Barough, M., Shirmardi, S.P.   +2 more
openaire   +1 more source

Comparison of GEANT4 and MCNPX-PoliMi fission models

2012 IEEE Nuclear Science Symposium and Medical Imaging Conference Record (NSS/MIC), 2012
The detection and characterization of special nuclear material (SNM) is of significant importance for nonproliferation applications. For neutron energies
Syed F. Naeem, Shaun D. Clarke, Sara A. Pozzi   +2 more
openaire   +1 more source

SOFT-RT: Software for IMRT simulations based on MCNPx code

Applied Radiation and Isotopes, 2016
Intensity Modulated Radiation Therapy (IMRT) is an advanced treatment technique, widely used in external radiotherapy. This paper presents the SOFT-RT which allows the simulation of an entire IMRT treatment protocol. The SOFT-RT performs a full three-dimensional renderization of a set of patient images, including the definitions of region of interest ...
Telma Cristina Ferreira, Fonseca, Tarcisio Passos Ribeiro, Campos   +1 more
openaire   +2 more sources

MCNPX evaluation of gamma spectrometry results in high radon concentration areas

Radiation Protection Dosimetry, 2014
The radon concentration in underground workplaces may reach tens of thousands of Bq m(-3). A simple MCNPXTM Monte Carlo (MC) model of a cave was developed to estimate the influence of radon on the in situ gamma spectrometry results in various geometries and radon concentrations.
L, Thinová, J, Solc
openaire   +2 more sources

MCNPX–BUCAL1 code to code verification through burnup analysis

Annals of Nuclear Energy, 2013
Abstract The availability of accurate burnup data is an essential first step in any systematic approach to enhancement of economics, safety and performance of a research reactor. This first step requires the utilization of a well verified burnup code system. In this work a newly home-developed burnup code called BUCAL1 is presented. The code provides
B. El Bakkari, T. El Bardouni, B. Nacir, C. El Younoussi, Y. Boulaich, H. Boukhal   +5 more
openaire   +1 more source

Characteristic evaluation of photoneutron in radiotherapy room using MCNPX

Journal of Instrumentation, 2015
Linear accelerators are now playing a pivotal role in radiotherapy and high energy photon beams of a strength exceeding 8 MV have recently been mainly used. However, when using high energy photons, neutron contamination due to photonuclear reaction develops.
E.-T. Park, J.-H. Kim, C.-S. Kim, S.-S. Kang   +3 more
openaire   +1 more source

MCNPX internal dosimetry studies based on the NORMAN-05 voxel model

Radiation Protection Dosimetry, 2007
Anthropomorphic computational models coupled with radiation transport codes are valuable tools in radiation protection dosimetry. In particular, they are very reliable for the estimate of the energy absorbed by different organs due to an incorporated radionuclide.
P, Ferrari, G, Gualdrini
openaire   +2 more sources

VVER-440 fresh core depletion benchmark analysis using MCNPX

International Journal of Nuclear Energy Science and Technology, 2020
The paper presents outcomes obtained by reproducing a VVER-440 lattice physics benchmark using the MCNPX code. A whole study is performed using different radial and axial regions within the whole core. Neutronics and thermal-hydraulics core design basis limits for the first-cycle are studied herein.
openaire   +1 more source

MCNPX and GEANT4 simulation of γ-ray polymeric shields

Pramana, 2015
In this work, the shielding ability of a polymeric compound with gadolinium for gamma radiation has been investigated. The conceptual calculation of radiation attenuation and energy absorption as a function of different Gd percentages and the calculation of total compound density are performed using MCNP and GEANT4.
openaire   +1 more source