A binary formatwith lists of particle state information, for interchanging particles between variousMonte Carlo simulation applications, is presented. Portable C code for file manipulation is made available to the scientific community, along with converters and plugins for several popular simulation packages. Program summary Program Title:MCPL Program Files doi: http://dx.doi.org/10.17632/cby92...

This study intends to predict the scope of neutron production and induced radioactivity for two therapy accelerators in Taiwan: 235 MeV proton and 400 MeV/A carbon ion accelerators. For radiation safety concern, much effort has been put into associated benchmark calculations to ensure the quality of predictions. Two Monte Carlo codes FLUKA and MCNPX were applied to calculate the neutron yield f...

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...

A new algorithm for energy-loss straggling in MCNP is demonstrated. An approximate but accurate energy-loss moment-preserving differential cross section is used in conjunction with single event Monte Carlo simulation through each condensed history step to show that highly accurate energy spectra, leakage currents, and dose profiles can be obtained. This new approach provides a viable and even p...

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 axial distribution of energy deposited in bremsstrahlung conversion targets is modelled using the Monte Carlo N -Particle (MCNP) code. Systems comprising a disc-shaped target, segmented axially, and a point source of paraxial, monoenergetic electrons (2, 7, 10, or 14 MeV) are considered. Thick tantalum and aluminium targets and also tantalum targets of optimum thickness are modelled. Energy...

background: obtaining high quality images in single photon emission tomography (spect) device  is the most important goal in nuclear medicine. because if image quality is low, the possibility of making a mistake in diagnosing and treating the patient will rise. studying effective factors in spatial resolution of imaging systems is thus deemed to be vital. one of the most important factors in sp...

The verification and validation (V&V) of the Serpent 2 Monte Carlo code for 3-D reactor dosimetry applications is being carried out at VTT. Two code-to-code computational benchmark cases were calculated by MCNP 2. efficiency different variance reduction techniques was appraised. part currently includes two experimental benchmarks from SINBAD database: Pool Critical Assembly-Pressure Vessel ...

Radiation Transport Simulation Studies Using MCNP for a Cow Phantom to Determine an Optimal Detector Configuration for a New Livestock Portal. (August 2012) Joe Justina, M.Sc., Mangalore University Co-Chairs of Advisory Committee: Dr. Craig M. Marianno Dr. Sunil S. Chirayath A large radiological accident will result in the contamination of surrounding people, animal, vegetation etc. In such a s...

Although the Monte Carlo method is considered to be the most accurate method available for solving radiation transport problems, its applicability is limited by its computational expense. Thus, biasing techniques, which require intuition, guesswork, and iterations involving manual adjustments, are employed to make reactor shielding calculations feasible. To overcome this difficulty, we have dev...