نتایج جستجو برای: mcnp code
تعداد نتایج: 168334 فیلتر نتایج به سال:
A binary formatwith lists of particle state information, for interchanging particles between variousMonte Carlo simulation applications, is presented. Portable C code for file manipulation is made available to the scientific community, along with converters and plugins for several popular simulation packages. Program summary Program Title:MCPL Program Files doi: http://dx.doi.org/10.17632/cby92...
in this study, a general theoretical model to foresee and calculate the current-voltagecharacteristics in a plateau zone and the associated efficiency of the neutron detector (sensitivity) is presented. this study is to complete the previous studies in this field. the model also considers electric field distortion resulting from the charge collection effect. the characteristics curve is obtaine...
The Neutron induced activity and decay heat are very important for the nuclear design of fusion reactor. In this paper, activation analysis of DEMO reactor with helium-cooled solid blanket (HCSB-DEMO) was performed. Ceramics Li4SiO4 and Beryllium options were considered as tritium breeding material and neutron multiplier, respectively. Chinese low-activation ferritic steel (CLF-1) developed in ...
Introduction: In this paper, by complete definition of human eye containing the various parts and their materials, the difference between this model and a homogeneous water phantom are compared for two ophthalmic plaques using 125I and 103Pd. Material and methods: The simulation of the two phantoms were performed in the MCNP-4C code and by using the geometry of a three-dimensional eye, differen...
Nowadays medical imaging systems like PET/CT are widely used in cancer diagnosis. Due to serious and irreversible harms of ionization radiations, protection of all those who are exposed is the main concern of health issues. The main basis of the calculation of the shielding design in the medical imaging systems is that the absorbed dose should not exceed the allowed limit. Using the Monte Carl...
In this work, the Isfahan Miniature Neutron Source Reactor (MNSR) is first simulated using the WIMSD code, and its fuel burn-up after 7 years of operation ( when the reactor was revived by adding a 1.5 mm thick beryllium shim plate to the top of its core) and also after 14 years of operation (total operation time of the reactor) is calculated. The reactor is then simulated using the MCNP code,...
This work illustrates a methodology based on photon interrogation and coincidence counting for determining the characteristics of fissile material. The feasibility of the proposed methods was demonstrated using the Monte Carlo-based MCNPX/MCNPPoliMi code system capable of simulating the full statistics of the neutron and photon field generated by the photon interrogation of fissile and non-fiss...
The MCNP code is a general Monte Carlo N-Particle Transport program that is widely used in health physics, medical physics and nuclear engineering for problems involving neutron, photon and electron transport. However, due to the stochastic nature of the algorithms employed to solve the Boltzmann transport equation, MCNP generally exhibits a slow rate of convergence. In fact, engineers and scie...
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