Results 181 to 190 of about 4,949 (201)

Comparisons of the Calculations Using Different Codes Implemented in MCNPX Monte Carlo Transport Code for Accelerator Driven System Target

Fusion Science and Technology, 2012
The MCNPX code offers options based on physics packages; the Bertini, ISABEL, INCL4 intra-nuclear models, and Dresner, ABLA evaporation-fission models and CEM2k cascade-exciton model. The study analyzes the main quantities determining ADS performance such as neutron yield, neutron leakage spectra, and neutron and proton spectra in the target and in the
ŞAHİN, SÜMER, Celik, Yurdunaz, Gunay, Mehtap, Sarer, Basar   +3 more
openaire   +2 more sources

Neutronic analysis and validation of boiling water reactor core designed by MCNPX code

Annals of Nuclear Energy, 2015
Abstract This paper presents a design of boiling water reactor BWR model using MCNPX to develop benchmarks for checking the fuel management computer code packages. MCNPX code based on Monte Carlo method, is used to design a three dimensional model for BWR fuel assembly in typical operating temperature and pressure conditions. A test case was compared
Ahmed Abdelghafar Galahom, I.I. Bashter, Moustafa Aziz   +2 more
openaire   +1 more source

SPENT FUEL POOL CRITICALITY ANALYSIS THROUGH MCNPX 2.6.0 CODE

Anais da Semana Nacional de Engenharia Nuclear e da Energia e Ciências das Radiações, 2022
Jean A. D. Salomé, Clarysson A. M. da Silva, Raphael H. M. Silva, Cláubia Pereira   +3 more
openaire   +1 more source

Modelling radiation transport in voxel geometries with the Monte Carlo code MCNPX

2004
The paper summarizes the research activities at ENEA Bologna to develop a voxel option within the Monte Carlo code MCNPX. An integrated methodology (MCNPX-VOXEL) has been presented that solves the radiation transport problem in voxel geometries and is based on an extension of the Monte Carlo Code MCNPX. A set of medical data (e.g.
G. Gualdrini, Daffara, Claudia, K. W. Burn, M. Pierantoni, P. Ferrari   +4 more
openaire   +1 more source

Calculation of the gamma radiation shielding efficiency of cement-bitumen portable container using MCNPX code

Progress in Nuclear Energy, 2021
Sonia M Reda, Hosam M Saleh
exaly  

Radiation shielding features using MCNPX code and mechanical properties of the PbO Na2O B2O3CaO Al2O3SiO2 glass systems

Composites Part B: Engineering, 2019
Shams A M Issa, Yasser B Saddeek, M I Sayyed   +2 more
exaly  

Investigation of photon shielding performances of some selected alloys by experimental data, theoretical and MCNPX code in the energy range of 81 keV–1333 keV

Journal of Alloys and Compounds, 2019
Ferdi Akman, M I Sayyed, Mrkkacal
exaly  

Radiation shielding and mechanical properties of Al2O3-Na2O-B2O3-Bi2O3 glasses using MCNPX Monte Carlo code

Materials Chemistry and Physics, 2019
Shams A M Issa, Reda Elsaman, Ozge Kilicoglu   +2 more
exaly  

Gamma radiation shielding properties of the hematite-serpentine concrete blended with WO3 and Bi2O3 micro and nano particles using MCNPX code

Radiation Physics and Chemistry, 2018
M I Sayyed, Shams A M Issa
exaly  

Comparative investigations of gamma and neutron radiation shielding parameters for different borate and tellurite glass systems using WinXCom program and MCNPX code

Materials Chemistry and Physics, 2018
M I Sayyed, Lakshminarayana Gandham
exaly