Results 111 to 120 of about 10,040 (154)

Dose distribution calculation with MCNP code in a research irradiator

Radiation Physics and Chemistry, 2020
Abstract 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. We have characterized the 3 fields of the irradiator to maximize the volume of irradiation. Percentage discrepancies below 15%,
B. Leal-Acevedo, I. Gamboa-deBuen
openaire   +3 more sources

MCNP-PoliMi: a Monte-Carlo code for correlation measurements

Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 2003
AbstractThe Monte-Carlo simulation of correlation measurements that rely on the detection of fast neutrons and photonsfrom fission requires that particle interactions in each history be described as closely as possible. The MCNP-PoliMi 1 code has been developed from the standard MCNP code to simulate each neutron–nucleus interaction as closely ...
S. A. POZZI, PADOVANI, ENRICO, MARSEGUERRA, MARZIO   +2 more
openaire   +2 more sources

Validation of the Monte Carlo code MCNP-DSP

Annals of Nuclear Energy, 1997
Abstract Several calculations were performed to validate MCNP-DSP, which is a Monte Carlo code that calculates all the time and frequency analysis parameters associated with the 252 Cf-source-driven time and frequency analysis method. The frequency analysis parameters are obtained in two ways: directly by Fourier transforming the detector responses ...
T.E. Valentine, J.T. Mihalczo
openaire   +1 more source

Comparison Between Anthropomorphic Mathematical Phantoms Using MCNP and FLUKA Codes

Radiation Protection Dosimetry, 1996
The International Commission on Radiation Protection (ICRP) recommends dose limits for occupational and public exposures in terms of weighted averages of organ and tissue doses. To this end the ICRP introduced the quantity Effective Dose Equivalent in Publication 26, substituted by Effective Dose in Publication 60.
M. Pelliccioni, M. Pillon
openaire   +1 more source

Benchmark study of OPAL research reactor using MCNP codes

Kerntechnik, 2019
Abstract This work is a part of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) titled: \Benchmarks of Computational Tools against Experimental Data on Fuel Burnup and Material Activation for Utilization, Operation and Safety Analysis of Research Reactors".
N. M. A. Mohamed, I. Abdelrazek, A. Badawi   +2 more
openaire   +1 more source

Using MCNP Code for Neutron and Photon Skyshine Analysis

Journal of Nuclear Science and Technology, 2000
The MCNP Monte-Carlo code was used for the investigation of the sensitivity of neutron and neutron-induced secondary photon dose rate, total and thermal neutron fluxes and space-energy distribution...
V P Zharkov, O F Dikareva, I A Kartashev, A N Kiselev, M E Netecha, Y Nomura, A Tsubosaka   +6 more
openaire   +1 more source

The development of an MCNP tally-based burnup code

International Journal of Nuclear Energy Science and Technology, 2009
The aim of this study is to evaluate the potential capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5 through the use of MCNP tally information. BUCAL1 uses the fourth-order Rung Kutta method with the predictor-corrector approach as an integration method to determine fuel ...
Bilal El Bakkari, Tarek El Bardouni, Ossama Merroun, Chafik El Younoussi, Yassine Boulaich, El Mahjoub Chakir   +5 more
openaire   +1 more source

Use of the MCNP-polimi code for time-correlation safeguards measurements

Progress in Nuclear Energy, 2003
Abstract Active nuclear safeguards measurements that rely on the time correlation between fast neutrons and gamma rays from the same fission are becoming a useful technique. In previous works we have shown the feasibility of this method, in conjunction with the use of the well-known MCNP simulation code and the use of artificial neural networks, to ...
MARSEGUERRA, MARZIO, PADOVANI, ENRICO, S. Pozzi   +2 more
openaire   +2 more sources

Development of the CAD/MCNP Automatic Conversion Code GEOMIT

Volume 2: Structural Integrity; Safety and Security; Advanced Applications of Nuclear Technology; Balance of Plant for Nuclear Applications, 2009
GEOMIT is the CAD/MCNP conversion interface code. The old version of GEOMIT had a limited capability from CAD model handling point of view. It is developed to automatically generate Monte Carlo geometrical data from CAD data due to the difference in the representation scheme.
Hesham R. Nasif, Fukuzo Masuda, Hiromasa Iida, Hidetsugu Morota, Satoshi Sato, Chikara Konno   +5 more
openaire   +1 more source

Criticality safety analysis using continuous energy libraries of MCNP code

International Journal of Nuclear Energy Science and Technology, 2015
The study of subcritical and critical systems is useful to guide decisions about design, installation and operation of some devices, mainly nuclear reactors and power plants. The information generated by these systems guides the best decisions to be made in executive project, economic viability and the safety measures to be employed in a nuclear ...
Jean A.D. Salomé, Claubia Pereira, Jonathan B.A. Assunsio   +2 more
openaire   +1 more source