نتایج جستجو برای: mcnpx code
تعداد نتایج: 168222 فیلتر نتایج به سال:
Neutron coincidence counting is a technique widely used for qualitative and quantitative analysis of nuclear material. Because different isotopes possess different coincident neutron characteristics, the coincident neutron signature can be used to identify and quantify a given material. In an effort to identify unknown nuclear samples in field inspections, a new portable neutron coincidence cou...
In this study, distribution of dose rate around the nuclear gauge device MC-1DR which located in shahrekord university was simulated by MCNPX code and was compared whit the measured values. Due to the asymmetry of device and neutron and gamma source positions, the dose rates were determined at a distance of 5, 30 and 100 cm in different directions. Base on the complex geometry of the inside of ...
In this research, water equivalent ratios (WER) at helium ion beam energy ranging 25-250 MeV/u for four potential plastic dosimetric materials: polycarbonate (PC), polypropylene (PP), polymethyl methacrylate (PMMA), and paraffin have been calculated using MCNPX Monte Carlo code. Among studied materials, PC and PP with 0.979 and 1.177 show minimum and maximum differences with water respectively....
Differential Die-Away Analysis (DDAA) is a very effective technique for detection of special nuclear material (SNM). It is based on fission neutrons that are detected with a time constant characteristic of the thermal neutron that created them. The presence of fast neutrons with this slow die-away time is a positive indication of the presence of SNM in the inspected cargo. Results are presented...
An epithermal neutron beam has been designed for Boron neutron Capture Therapy (BNCT) at the thermal column of Tehran Research Reactor (TRR) recently. In this paper the whole body effective dose, as well as the equivalent doses of several organs have been calculated in this facility using MCNPX Monte Carlo code. The effective dose has been calculated by using the absorbed doses determined for e...
This study presents neutronic analyses of a small modular reactor utilizing transuranium and thorium. Two different fuel cases are considered in the as extracted from PWR-MOX spent (a form mixture minor actinide Pu isotopes) (Case A) 4.5 % enriched UO2 with ThO 2(the separate rods) B). The total power containing 69 assemblies is 450 MW thermal. In both cases, time-dependent critical burnup calc...
Introduction Electron linear accelerator (LINAC) can be used for neutron production in Boron Neutron Capture Therapy (BNCT). BNCT is an external radiotherapeutic method for the treatment of some cancers. In this study, Varian 2300 C/D LINAC was simulated as an electron accelerator-based photoneutron source to provide a suitable neutron flux for BNCT. Materials and Methods Photoneutron sources w...
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