نتایج جستجو برای: mcnp calculation
تعداد نتایج: 102713 فیلتر نتایج به سال:
In this work we present a predictive analytical model that encompasses the performance and scaling characteristics of a non-deterministic particle transport application, MCNP. Previous studies on the scalability of parallel Monte Carlo eigenvalue calculations have been rather general in nature [1]. It can be used for the simulation of neutron, photon, electron, or coupled transport, and has fou...
We present new results on neutron and gamma-ray pulse-height distributions (PHDs) measured with liquid scintillators from five plutonium-oxide samples of varying mass and burnup and a Cf isotopic source. We show that the analysis of the pulse-height distributions can be used to easily distinguish the fissile material (plutonium oxide) from the Cf source. Moreover, the slope of the measured puls...
Background: Monte Carlo and experimental relative dose determination in a water phantom, due to a high dose rate (HDR) 192Ir source is presented for real energy spectrum and monochromatic at 356 keV. Materials and Methods: The dose distribution has been calculated around the 192Ir located in the center of 30 cm ×30 cm ×30 cm water phantom using MCNP4C code by Monte Carlo method. Relati...
In the recent years some studies has been done to consider the capability of Tehran Research Reactor for Boron neutron capture therapy (BNCT). The purpose of this study is to evaluate the sensitive organs dose during the treatment of patient with deep brain tumor by TRR. The calculation has been carried out using the Monte Carlo code MCNPX for ZUBAL head and neck phantom. The method was tested ...
The modelling with the Geant4 toolkit of moderated 3He-filled proportional neutron counters is studied. The energy deposition spectra by the neutron capture products in the gas counter is compared with experimental results. On the other hand, efficiency calculation for polyethylene moderated proportional counters are compared with experimental and MCNPX results.Finally, the application of the G...
Neutron detection time distribution is an important factor for the dead-time correction for moderator type neutron detectors used in pulsed radiation fields. Measurements of the neutron detection time distributions of multisphere LiI detectors (2Ó, 3Ó, 5Ó, 8Ó, 10Ó and 12Ó in diameter) and AB remmeter were made inside a ANL 20-MeV electron Linac room. Calculations of the neutron detection time d...
Neutron transport through a special case stochastic mixture is examined, in which spheres of constant radius are uniformly mixed in a matrix material. A Monte Carlo algorithm previously proposed and examined in 2-D has been implemented in a test version of MCNP. The Limited Chord Length Sampling (LCLS) technique provides a means for modeling a binary stochastic mixture as a cell in MCNP. When i...
BACKGROUND PET/CT has recently been shown to be a viable alternative to traditional post-infusion imaging methods providing good quality images of 90Y-laden microspheres after selective internal radiation therapy (SIRT). In the present paper, first we assessed the quantitative accuracy of 90Y-PET using an anthropomorphic phantom provided with lungs, liver, spine, and a cylindrical homemade lesi...
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