Results 181 to 190 of about 4,292 (208)
Some of the next articles are maybe not open access.
MCNPX vs. DORT for sns shielding design studies
Radiation Protection Dosimetry, 2005Radiation transport occurs through the 18 m long access way adjacent to the Spallation Neutron Source accelerator tunnel and the 2.2 m thick massive shielding door which closes the access way. A variety of typical materials for accelerator shielding, such as concrete and steel, were used for construction of the door to study radiation penetration.
openaire +2 more sources
Comparison of GEANT4 and MCNPX-PoliMi fission models
2012 IEEE Nuclear Science Symposium and Medical Imaging Conference Record (NSS/MIC), 2012The detection and characterization of special nuclear material (SNM) is of significant importance for nonproliferation applications. For neutron energies
Syed F. Naeem +2 more
openaire +1 more source
Prediction of 67Ga production using the Monte Carlo code MCNPX
Applied Radiation and Isotopes, 2013The widely used Monte Carlo simulation code Monte Carlo N-Particle System (MCNPX) has been utilized to simulate the production of (67)Gallium via multiple nuclear reaction channels. Based on the MCNPX-generated, energy-dependent proton flux within a Zn target during irradiation, the (67)Ga production yield was determined.
M, Sadeghi +4 more
openaire +2 more sources
Computed radiography simulation using the Monte Carlo code MCNPX
Applied Radiation and Isotopes, 2010Simulating X-ray images has been of great interest in recent years as it makes possible an analysis of how X-ray images are affected owing to relevant operating parameters. In this paper, a procedure for simulating computed radiographic images using the Monte Carlo code MCNPX is proposed.
S C A, Correa +4 more
openaire +2 more sources
SOFT-RT: Software for IMRT simulations based on MCNPx code
Applied Radiation and Isotopes, 2016Intensity 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 +1 more
openaire +2 more sources
MCNPX calculation of the reactivity worth of actinides OSMOSE samples
Annals of Nuclear Energy, 2015Abstract Improving neutronic prediction is a very important step in designing advanced reactors and reactor fuel. There are three main critical reactor facilities of the CEA Cadarache. These reactor facilities are EOLE, MINERVE and MASURCA. The MINERVE reactor is used within the frame work of what is known as OSMOSE project.
Hesham I. Shahbunder +3 more
openaire +1 more source
Calculation of remanent dose distributions with MCNPX
2007A script was developed that automates the extraction of gamma source spectra from the buildup and decay results of the codes CINDER’90 or SP-FISPACT, and then generates a source description for subsequent gamma transport analysis with MCNP(X). This new script was used to calculate remanent dose rates from activated components in accelerator facilities.
Wohlmuther, M. +5 more
openaire +1 more source
MCNPX and GEANT4 simulation of γ-ray polymeric shields
Pramana, 2015In 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
Gamma photon-neutron attenuation parameters of marble concrete by MCNPX code
Radiation Effects and Defects in Solids, 2021İskender Akkurt +1 more
exaly
Comparison of MCNPX and FLUKA Monte Carlo codes in the simulating a nuclear gauge
Applied Radiation and Isotopes, 2021Mojtaba Tajik
exaly