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...

Pohang Neutron Facility (PNF) is a pulsed neutron facility based on the 100-MeV electron linear accelerator. It was constructed for nuclear data production in Korea, and it consists of an electron linear accelerator, a watercooled Ta target with a water moderator and a time-offlight path with an 11 m length. The 100-MeV electron linac uses a thermionic RF-gun, an alpha magnet, four quadrupole m...

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...

The csg2csg tool is a minimal dependency Python3 based program to facilitate the comparison of radiation transport Monte Carlo codes. can currently parse MCNP and OpenMC geometry descriptions, it subsequently export MCNP, FLUKA, PHITS, Serpent files. This translation achieved by reading input into class system objects corresponding surfaces, cells, materials so on. are then translated ‘generic’...