The Use of the MCNP Code for Radiation Damage Calculations

This work gives a detailed analysis of the result of Monte Carlo physics practical using MCNP. This paper describes basic concepts of the Monte Carlo theory of radiation transport calculation and also discusses the variance and the history method as used in Monte Carlo Problem solving. Therefore, in this exercise the MCNP code has been used to solve and estimate the number of neutron flux. The paper investigated the impact of the primary radiation damage in iron by the neutron energy irradiation. The established measurement of radiation damage is the displacements per atom (dpa) in matter as a function of neutron energy. The simulations were carried out to calculate the dpa cross section.

Radiological Characterization of Contaminated Pipes of Different Dimensions and Densities

The present work concerns a preliminary study for development of a technique for radiological characterization and segregation of raw historical radioactive waste in different management routes. The efficiency of a 3x3 NaI (Tl) detector for a contaminated cylindrical pipe - detector configuration was evaluated by Monte Carlo simulations performed by using the MCNP code. The efficiency for detector source configuration as well as the measurement bias due to possible inhomogeneity in the distribution of the activity were examined for cylindrical pipes of different densities and dimensions. Cylinders of three different densities made by Pb, Fe, Al were examined. All studied cylinders keep the same ratio between length, diameter and thickness, although the absolute values are different, in order to study the difference in efficiencies of similar geometric objects

On the Rapid Generation of Complete XRF Spectra for Material Analysis from Fundamental Parameters

X-ray Fluorescence (XRF) analysis technology is used widely to detect and measure elemental compositions of target samples. The MCNP code developed by LANL can be utilized to simulate and generate the XRF spectrum of any sample with various elemental compositions. However, one shortcoming of MCNP code is that it takes quite a lot of time (in hours or longer) to generate one XRF spectrum with reasonable statistical precision; the other shortcoming is that MCNP code cannot produce L shell spectrum accurately. In this paper, a new computation model based on the Sherman equation (i.e., Fundamental Parameters, FP) is proposed to overcome the drawbacks of the MCNP code. The most important feature of this model is to achieve a full and accurate generation of spectral information of each element in a target material very rapidly (in seconds or less), including both K and L shell spectral peaks. Furtherly, it is demonstrated that the simulated data by this new mode match the experimental data very well. It proves that the proposed model can be a better alternative of MCNP code in the application of generation the XRF spectra of many materials, in terms of speed and accuracy. The proposed model can perform the simulation of XRF spectra in situ both fast and accurately, which is essential for real-time calculation of chemical composition by use of X-ray spectrometer, especially for those trace elements in target materials.

EGSnrc/BEAMnrc-Based Monte Carlo Simulation of the Gamma Knife 4C versus MCNP Code in Homogeneous Media

Purpose: Gamma Knife is applied as a superseded tool for inaccessible lesions surgery delivering a single high dose to a well-defined target through 201 small beams. Monte Carlo simulations can be an appropriate supplementary tool to determine dosimetric parameters in small fields due to the related dosimetry hardships. Materials and Methods: EGSnrc/BEAMnrc Monte Carlo code was implemented to model Gamma Knife 4C. Single channel geometry comprising stationary and helmet collimators was simulated. A point source was considered as a cylindrical Cobalt source based on the simplified source channel mode. All of the 201 source channels were arranged in spherical coordinate by EGSnrc/DOSXYZnrc code to calculate dose profiles. The simulated profiles at the isocentre point in a spherical head phantom 160 mm in diameter along three axes for 4, 8, 14, and 18 mm field sizes were compared to those obtained by another work using MCNP code. Results: Based on the results, the BEAMnrc and MCNP dose profiles matched well apart from the 18 mm profiles along X and Y directions with the average gamma index of 1.36 and 1.18, respectively. BEAMnrc profiles for 14 and 18 mm field sizes along X and Y axes were entirely flat in plateau region, whereas MCNP profiles represented variations as well as round shape. Besides, considering the identical results, radioactive source can be modeled by a point source instead of cylindrical one. Conclusion: Thus, the EGSnrc/BEAMnrc code is recommended to simulate Gamma Knife machine as it is regarded as the most accurate computer program to simulate photon and electron interactions.