Results 161 to 170 of about 5,690 (213)

Simulation of neutron production using MCNPX+MCUNED

Radiation Protection Dosimetry, 2013
In standard MCNPX, the production of neutrons by ions cannot be modelled efficiently. The MCUNED patch applied to MCNPX 2.7.0 allows to model the production of neutrons by light ions down to energies of a few kiloelectron volts. This is crucial for the simulation of neutron reference fields.
M, Erhard, P, Sauvan, R, Nolte
openaire   +2 more sources

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

Materials Chemistry and Physics, 2019
The glasses had the compositions 5Al2O3-29Na2O-(66−x)B2O3-xBi2O3 with (0 ≤ x ≤ 20 mol%) are prepared by melt quenching technique for serving as competitive candidates for radiation shielding.
S. Issa, H. O. Tekin, R. Elsaman, O. Kilicoglu, Y. Saddeek, M. Sayyed   +5 more
semanticscholar   +1 more source

Delayed-Gamma Simulation Using MCNPX

Nuclear Technology, 2009
Monitoring issues related to activation and fission processes occur in many health physics, instrumentation and equipment design, nuclear forensics, and homeland security applications. Gamma radiation that is emitted during these processes as a result of the radioactive decay of reaction by-products [delayed gammas (DGs)] provides unique signatures ...
Joe W. Durkee, Gregg W. McKinney, Holly R. Trellue, Laurie S. Waters, William B. Wilson   +4 more
openaire   +1 more source

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
Recently, due to the increasing applications of nano materials, among researchers, there has been great attention paid to the enhancement of the conventional concrete properties.
H O Tekin, M I Sayyed, Shams A M Issa
exaly   +2 more sources

An antiproton simulation study using MCNPX for radiation therapy

Journal of X-Ray Science and Technology: Clinical Applications of Diagnosis and Therapeutics, 2011
Radiation therapy using antiprotons is a potential interesting future modality. Energetic antiprotons penetrate matter with almost near identical stopping powers and radio biological effectiveness (RBE) as protons in the region well before the Bragg peak region.
Stephen, Michael Handley, Salahuddin, Ahmad   +1 more
openaire   +2 more sources

Non-static surfaces in MCNPX: The chopper extension1

Journal of Neutron Research, 2020
Rotating objects, such as choppers, are common components of a neutron beamline, and the motion of these components is not described in the static geometry of an MCNPX model. The special case of non-static surfaces for rotation about a stationary point in space has been developed for MCNPX.
Grammer, Kyle B., Gallmeier, Franz X., Iverson, Erik B.   +2 more
openaire   +2 more sources

Radiation shielding study of tellurite tungsten glasses with different antimony oxide as transparent shielding materials using MCNPX code

Journal of Non-Crystalline Solids, 2018
The radiation shielding performance of tellurite tungsten glasses with different antimony oxide (Sb2O3) content was studied using Monte Carlo simulation MCNPX and WinXCOM software.
M I Sayyed, H O Tekin, Shamsan S Obaid
exaly   +2 more sources

A new implementation of scattering kernels in MCNPX

Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 2004
Abstract With the increased demand for cold neutrons the designers of cold sources are challenged to improve the quality of theoretical predictions of the behavior of such sources. This challenge can be met by improving codes like MCNPX or by improving the physics behind the codes. This paper discusses the impact improving the code has on the results
G Muhrer, A Stupnik, E Schachinger
openaire   +1 more source

Full core analysis of IRIS reactor by using MCNPX

Applied Radiation and Isotopes, 2016
This paper describes neutronic analysis for fresh fuelled IRIS (International Reactor Innovative and Secure) reactor by MCNPX code. The analysis included criticality calculations, radial power and axial power distribution, nuclear peaking factor and axial offset percent at the beginning of fuel cycle.
E A, Amin, I I, Bashter, Nabil M, Hassan, S S, Mustafa   +3 more
openaire   +2 more sources

Modeling the thermoluminescent response of CaSO4:Dy by the MCNPX method

Applied Radiation and Isotopes, 2010
This work describes an algorithm for absorbed dose evaluation in the region of X-ray diagnostic energy based on the response of CaSO(4):Dy thermoluminescence powder detectors. The absorbed dose was calculated using Monte Carlo simulation code (MCSC) and then compared to the experimental TLD results for X-rays effective energies (60)Co gamma radiation ...
C, Furetta, J, Roman, T, Rivera, J, Azorín, C G, Azorín, H R, Vega-Carrillo   +5 more
openaire   +2 more sources