Results 161 to 170 of about 4,949 (201)

Gamma photon-neutron attenuation parameters of marble concrete by MCNPX code

Radiation Effects and Defects in Solids, 2021
The present study aims to investigate the gamma photon-neutron attenuation parameters of the marble concrete such as the linear attenuation coefficient (mu), Mass Attenuation Coefficient (MAC), Half Value Layer (HVL), neutron removal cross-section (Sigma(R)), and Exposure Buildup Factor (EBF) by MCNPX Monte Carlo code to examine the possibility of ...
AKKURT, İskender, Boodaghi Malidarre, Roya   +1 more
openaire   +5 more sources

Radiation shielding properties of pentaternary borate glasses using MCNPX code

Journal of Physics and Chemistry of Solids, 2018
Abstract The mass attenuation coefficient, effective atomic number, and half value layer of (75-x) B 2 O 3 -xBi 2 O 3 -10Na 2 O-10CaO-5Al 2 O 3 glass system (x = 0, 5, 10, 15, 20 and 25mol%) with potential applications as radiation shielding materials, have been investigated in the energy ranging from 31 keV to 662 keV using the MCNPX Code ...
M.I. Sayyed, B.O. Elbashir, H.O. Tekin, E.E. Altunsoy, D.K. Gaikwad   +4 more
openaire   +3 more sources

Modeling NE213 scintillator response to neutrons using an MCNPX-PHOTRACK hybrid code

Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 2013
This paper reports on how to generate the response function of an NE213 scintillator when exposed to mono-energetic neutrons using the PTRAC card of the MCNPX code. The light transport part of the simulation has been undertaken with the Monte Carlo PHOTRACK code.
M. Tajik, N. Ghal-Eh, G.R. Etaati, H. Afarideh   +3 more
openaire   +3 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

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

Computed radiography simulation using the Monte Carlo code MCNPX

Applied Radiation and Isotopes, 2010
Simulating 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, E M, Souza, A X, Silva, D H, Cassiano, R T, Lopes   +4 more
openaire   +2 more sources

Comparison of MCNPX and FLUKA Monte Carlo codes in the simulating a nuclear gauge

Applied Radiation and Isotopes, 2021
In this paper, a nuclear gauge was simulated using MCNPX and FLUKA codes. This device consists of a fluid-containing vessel, three BC400 rod plastic scintillator and a60Co gamma source. Simulation studies show that the changes in the count of each of the three detectors and the sum of their counts decrease with increasing vessel water height.
Saeid, Mohtaram, Mojtaba, Tajik, Reza Gholipour, Peyvandi   +2 more
openaire   +2 more sources

Photonuclear physics modeling in the MCNPX-PoliMi code

2012 IEEE Nuclear Science Symposium and Medical Imaging Conference Record (NSS/MIC), 2012
The MCNP-PoliMi code is an enhanced version of the MCNP4c code developed in 2003 to simulate the detector response from correlated neutron and gamma-ray measurements. In 2012 the -PoliMi modifications were incorporated into version 2.7.0 of the MCNPX code.
Shaun D. Clarke, Sara A. Pozzi, Enrico Padovani   +2 more
openaire   +1 more source

Prediction of 67Ga production using the Monte Carlo code MCNPX

Applied Radiation and Isotopes, 2013
The 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, N, Jokar, T, Kakavand, H, Ghafoori Fard, C, Tenreiro   +4 more
openaire   +2 more sources

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