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...

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 MCNP Monte Carlo radiation transport code was used for calculation of scatter distribution and scatter-toprimary ratio (SPR) for different tube voltages, phantom thicknesses and field size in diagnostic radiology. Subsequently, the effect of grid’s design parameters such as strip density, grid ratio, interspace material and lead-tointerspace ratio on scatter rejection was investigated by ca...

 Introduction: Monte Carlo method is often applied in radiation therapy as utilized in all the branches of science. An important requirement for successful radiotherapy is carefully examine the dose distribution specifications and decrease the difference between these features with experience to an acceptable level. In this study, the characteristics of 6, 15 and 20 MeV incident x-rays are prov...

The Monte Carlo calculation of the effective (i.e. adjoint weighted) neutron generation time Keff, especially for continuous energy simulations, is not straightforward nor standard in Monte Carlo codes. The use of the non-adjoint weighted neutron generation time K (standard in most Monte Carlo codes) as an approximation of the effective one can lead to a serious bias. We show here that the diff...

The ability to quickly quantify the Pu content within spent nuclear fuel (SNF) is essential to nuclear forensics. Analysis of the Pu to U ratio can provide information on fuel which could contribute to the attribution of a fuel sample. Plutonium concentration data can be acquired through non-destructive analysis (NDA) by detecting self-induced x-ray fluorescence (XRF) from Pu in the fuel. Howev...

background: monte carlo and experimental relative dose determination in a water phantom, due to a high dose rate (hdr) 192ir source is presented for real energy spectrum and monochromatic at 356 kev. materials and methods: the dose distribution has been calculated around the 192ir located in the center of 30 cm ×30 cm ×30 cm water phantom using mcnp4c code by monte carlo method. relative dose v...

The aim of this study is the development of a methodology to reconstruct the image of detail defects from a mammographic phantom employed in quality control testing. The MCNP-4c2 code has been used to model the mammography unit and the CIRS 11A (MAMMO PHANTOM SP01) mammographic phantom. This phantom is made of poly-methyl methacrilate and contains a reference point, contrast and resolution deta...

Anatomically accurate phantoms are useful tools for radiation dosimetry studies. In this work, we demonstrate the construction of a new generation of life-like mouse phantoms in which the methods have been generalized to be applicable to the fabrication of any small animal. The mouse phantoms, with built-in density inhomogeneity, exhibit different scattering behavior dependent on where the radi...