Double heterogeneity of randomly located pebbles in the core and Coated Fuel Particles (CFPs) in the pebbles are specific features in pebble bed reactors and usually, because of difficulty to model with MCNP code capabilities, are neglected. In this study, characteristics of HTR-10, Tsinghua University research reactor, are used and not only double heterogeneous but also truncated CFPs and Pebb...

A great challenge in simulations in the area of radiotherapy is to character the source, since the equipment’s manufacturers don’t provide specifics information about this. This work presents an empiric method for characterization of the ray-X beams of 6,0 MV and 10 MV originating from a linear accelerator of the mark Varian model 2100C. The experimental values of the percentage of deep dose (P...

The neutron energy spectra have been calculated in the fuel region, inner and outer irradiation sites of the zero power research reactor using the MCNP-4C code and the combination of the WIMS-D/4 transport code for generation of group constants and the three-dimensional CITATION diffusion code for core analysis calculations. The neutron energy spectrum has been divided into three regions and co...

The detailed China Fusion Engineering Test Reactor (CFETR) 22.5° computer aided design (CAD) model is very difficult to convert into Monte Carlo N Particle Transport Code (MCNP). Manually writing MCNP input data complicated, which not only time-consuming but also cannot guarantee accuracy. Therefore, in order improve the efficiency and accuracy of transformation, modeling with CAD using CATIA i...

MCNP is the Monte Carlo N-Particle radiation transport code whose history dates back more than half a century to the early days of computing. From a simple beginning, its uses have grown to include fields such as criticality safety, radiation shielding, oil well logging, and medical imaging and diagnostics and an international user community of over 3000 users. This large user community could o...

The manganese bath method for measuring the neutron emission rate of radionuclide sources requires corrections to be made for emitted neutrons which are not captured by manganese nuclei. The Monte Carlo particle transport code MCNP has been used to calculate the fractions of neutrons that escape from the boundaries of the tank, that are captured by the steel cavity in which the source is housed...

The REBUS-3 burnup code, used in the ANL RERTR Program, is a very general code that uses diffusion theory (DIF3D) to obtain the fluxes required for reactor burnup analyses. Diffusion theory works well for most reactors. However, to include the effects of exact geometry and strong absorbers that are difficult to model using diffusion theory, a Monte Carlo method is required. MCNP, “a general-pur...

The energy spectrum of the Neutron Radiation Effects Program (NREP) beam line, Target-ModeratorReflector-1 (TMR-1), at Indiana University has not been previously characterized. The facility has a unique proton source with variable pulse length (15-600 s) and energy (13 MeV). Thus, it can produce a unique and tailored neutron beam when incident on a beryllium target. Through a combination of MCN...

AIM Exact knowledge of dosimetric parameters is an essential pre-requisite of an effective treatment in radiotherapy. In order to fulfill this consideration, different techniques have been used, one of which is Monte Carlo simulation. MATERIALS AND METHODS This study used the MCNP-4C to simulate electron beams from Neptun 10 PC medical linear accelerator. Output factors for 6, 8 and 10 MeV el...