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...