INTRODUCTION Monte Carlo (MC) methods have become the "gold standard" for assessing dose distribution, organ doses and effective dose associated with the exposure, both external and internal, of humans to ionizing radiation. Powerful and sophisticated tools for simulation of radiation transport (for example, MCNP, EGS4) are now widely available. However, these codes are designed for general-pur...

introduction bnct is an effective method to destroy brain tumoral cells while sparing the healthy tissues. the recommended flux for epithermal neutrons is 109 n/cm2s, which has the most effectiveness on deep-seated tumors. in this paper, it is indicated that using d-t neutron source and optimizing of beam shaping assembly (bsa) leads to treating brain tumors in a reasonable time where all iaea ...

With the advent of inexpensive computing power over the past two decades, applications of Monte Carlo radiation transport techniques have proliferated dramatically. At Los Alamos, the Monte Carlo codes MCNP5 [1] and MCNPX [2] are used routinely on personal computer platforms for radiation shielding analysis and dosimetry calculations. These codes are attractive for such applications because of ...

Prediction of neutron flux at the irradiation devices of a research reactor facility is essential for the design and evaluation of experiments involving material irradiations. A computational model of the Greek Research Reactor (GRR-1) was developed using the Monte Carlo code MCNP with continuous energy neutron cross-section data evaluations from ENDF/B-VI library. The model included detailed g...

Series of mathematical phantoms of human body, given by Oak Ridge National Laboratory (ORNL), was programmed as input files for MCNP-4B code. Detailed check of geometry of these phantoms performed by MCNP-4B, discovered some minor errors, that resulted in overlapping of some organs. Three types of errors were found and described here: (a) colon overlaps with pelvis bone; (b) facial skeleton pen...

Abstract The contents and distribution of radionuclides in activated metallic wastes are major concerns while designing a leaching model safety case. This study investigated 60 Co as typical activation product the lower tie plate (i.e., stainless steel endpiece BWR fuel assembly) irradiated for 35.0 GWd/tHM burnup, using imaging (IP) method. calculations coupled with Monte Carlo burnup neutron ...

background: several landmine detection methods, based on nuclear techniques, have been suggested during the recent years. neutron energy moderation, neutron-induced gamma emission, neutron and gamma attenuation, and fast neutron backscattering are nuclear-based methods used for landmine detection. the aim of this study is to use backscattered neutron for landmine detection. materials and method...

background: monte carlo determination of tg-43 brachytherapy dosimetry parameters and dose distribution calculation for 131cs source model cs-1 are presented in this study. materials and methods: the dose distribution was calculated around the 131cs model cs-1 located in the center of 30 cm ×30 cm ×30 cm water, and soft tissue phantoms cube using mcnp code by monte carlo method. the percentage ...

Neutron dose-equivalent buildup factors were calculated for point isotropic neutron sources irradiating infinite slablike shields of lead, iron and water using the MCNP code. The factors are presented for some source neutron energies in the range from 0.025 eV to 14 MeV, and for shield thicknesses from 0.5 to 10 mfp. Since the MCNP code considers all kinds of neutron interactions with matter, a...