Results 191 to 200 of about 25,046 (241)

Application of the MCNP 5 code to simulate the shielding features of concrete samples with different aggregates

, 2020
The gamma ray attenuation parameters for three different series of concretes containing different ratios of basalt, magnetite, and bismite mineral aggregates were studied using the Monte Carlo simulation MCNP 5 code.
M. Sayyed, K. Mahmoud, Sazirul Islam, O. Tashlykov, E. Lacomme, Kawa M. Kaky   +5 more
semanticscholar   +1 more source

Validation of two calculation options built in Elekta Monaco Monte Carlo based algorithm using MCNP code

, 2021
Introduction The calculation algorithm for intensity modulated radiotherapy (IMRT) built in the Elekta Monaco treatment planning system (TPS) is based on Monte Carlo (MC) simulation. Absorbed dose is calculated as dose to medium in medium (Dm,m), but the
Manda Švabić Kolacio, H. Brkić, D. Faj, Đ. Radojčić, D. Rajlić, N. Obajdin, S. Jurković   +6 more
semanticscholar   +1 more source

MCNP Monte Carlo: a precis of MCNP

SPIE Proceedings, 1997
MCNPTM Monte Carlo: a precis of MCNPKenneth J. AdamsTransport Methods Group, XTM MS B226Los Alamos National Laboratory, Los Alamos, NM, 87545ABSTRACTMCNPTM* is a general purpose tiiree-dimensional time-dependent neutron, photon, and electron transport code.
openaire   +1 more source

Investigation of gamma-ray shielding capability of glasses doped with Y, Gd, Nd, Pr and Dy rare earth using MCNP-5 code

, 2020
The capability of using five rare earth oxides (Y2O3 (S1-Y), Gd2O3 (S2-Gd), NdCl3 (S3-Nd), PrO11 (S4-Pr), and Dy2O3 (S5-Dy)) doped glasses as gamma-rays shielding has been evaluated. The mass attenuation coefficient (MAC, μm) in the photon energy between
K. Mahmoud, Y. Rammah
semanticscholar   +1 more source

Comparison of gamma ray localization using system matrixes obtained by either MCNP simulations or ray-driven calculations for a coded-aperture imaging system

, 2020
A coded-aperture system based on a large-area silicon photo-multiplier (SiPM) coupled with inorganic scintillator was developed for the gamma ray localization in the field of nuclear safety and security.
M. Jeong, M. Hammig
semanticscholar   +1 more source

MCNP benchmark of a 252Cf source-based PGNAA system for bulk sample analysis.

Applied Radiation and Isotopes, 2020
A MCNP model of a252Cf source-based prompt gamma-ray neutron activation analysis (PGNAA) system for bulk sample analysis was benchmarked. The indium foils activation and aqueous samples containing chlorine were measured using the PGNAA system, and the ...
Can Cheng, Zhi-yong Wei, D. Hei, Wenbao Jia, Jiatong Li, Pingkun Cai, Youtao Gao, Q. Shan, Yongsheng Ling   +8 more
semanticscholar   +1 more source

The comparison of MCNP perturbation technique with MCNP difference method in critical calculation

Nuclear Engineering and Design, 2010
Abstract For a nuclear fission system, we calculated Δkeff, which arise from system material composition changes, by two different approaches, the MCNP perturbation technique and the MCNP difference method. For every material composition change, we made four different runs, each run with different cycles or each cycle generating different neutrons ...
Bin Liu, Xuefeng Lv, Wei Zhao, Kai Wang, Jing Tu, Xiaoping Ouyang   +5 more
openaire   +1 more source

Simulation studies for gamma ray shielding properties of Halloysite nanotubes using MCNP-5 code.

Applied Radiation and Isotopes, 2019
Halloysite clay is a mineral found in natural and it has many applications in chemistry (for catalytic and extraction) and also in the medical field (for drugs delivery), so it is important to study the shielding properties of natural and modified ...
A. M. Abu El-Soad, M. Sayyed, K. Mahmoud, E. Şakar, E. Kovaleva   +4 more
semanticscholar   +1 more source

Comparative studies between the shielding parameters of concretes with different additive aggregates using MCNP-5 simulation code

, 2019
In the present work, cementation concretes with different aggregates were produced. The mass attenuation coefficient ( μ m ) for the produced concretes is simulated using MCNP 5 code between 0.047 MeV and 1.836 MeV.
K. Mahmoud, M. Sayyed, O. Tashlykov
semanticscholar   +1 more source

Dose distribution calculation with MCNP code in a research irradiator

, 2020
By comparing the data obtained experimentally, the flux density calculations and simulation results using MCNP we can correlate all the results to ensure that the dose imparted to industrial products is appropriate.
B. Leal-Acevedo, I. Gamboa-deBuen
semanticscholar   +1 more source