Model based on a track structure Monte Carlo simulation of low-energy electron transport in thin film materials is presented. It is a modification of previous models of its kind, and it takes into account all relevant interactions elastic scattering, inelastic scattering and bremsstrahlung. Results obtained with the new model are compared to those from MCNP5 simulations, as well as to experimental data. It is demonstrated that the proposed model can be utilized for practical calculations. The model shows better agreement with experimental values of the backscatter coefficient than MCNP5, with no significant increase in computation time. Detailed Monte Carlo simulation of electron transport for electron’s energies up to40 keV Predrag Marinkoví c, Jelena Stankovi ć-Petroví c, Miloš Vujisić, Radovan Ilí c

Efficiency of a Monte Carlo algorithm for neutron dose calculation is compared in two implementations: a standard C++ code executed sequentially, and a CUDA C/C++ code which utilizes GPU resources for highly parallel processing. Both versions of the algorithm, developed specifically for this investigation, are based on the same physical model for the assessment of neutron dose in tissues, including lung, cortical bone and adipose tissue. The model treats emission and interaction of neutrons stochastically, utilizing cross sections for relevant interaction types. Several intentional simplifications have been introduced into the physical model used for simulations, which have allowed parts of the two codes to be related to one another in a straightforward way. A neutron's history is terminated when it leaves the outer ellipsoid (representing the human body), experiences any of the absorption interactions (inside one of the inner geometrical regions, representing tissues or organs), or if its energy falls below the cut-off limit set at 0.001 eV. The two approaches to algorithm implementation are compared according to execution speed, at various neutron source energies and for an increasing number of neutron histories. The fact that particle histories in a Monte Carlo simulation are independent from one another makes this kind of calculation suitable for implementation on parallel processing platforms. CUDA framework offers higher speeds of code execution, allowing more particle histories to be processed within a set time frame, and thus yields lower statistical uncertainty and higher reliability of the calculated neutron dose values. Appropriating standard C++ codes for CUDA is faced with specific challenges, which are described in the investigated case of neutron dose assessment. Despite the physical representation of neutron transport being somewhat simplified, comparison of both implementations to results obtained from MCNP shows good agreement in a wide range of neutron energies.

Pulse height distribution (PHD) registered by a spectrometer is influenced by various physical phenomena such as photon interactions as well as disturbance produced by the electronic circuits inside the spectrometer. Therefore, spectrometry measurements of gamma and X-ray radiation inaccurately represent primary spectra. In order to overcome spectrum disruption, spectrum unfolding has to be applied. One of the common tools used in the unfolding process is Monte Carlo simulation of spectrometer response to monochromatic photons. The purpose of this work is to develop a new method for simulating CdTe semiconductor spectrometer response to monochromatic photons that can be further used for the spectrum unfolding procedure. The method is based upon post-processing of the particle track (PTRAC) output file generated by the MCNP5 program. In addition to the spectrometry output, this method provides information for each specific photon interaction inside the spectrometer active volume, which is required when taking into account spectrometer charge collection. The PTRAC generated detector response and the measured spectrum were in good agreement. The results obtained showed that this method can be used to generate precise response functions of gamma and X-ray spectrometers.

This paper deals with the design of a novel spectrometer of fast neutrons in nuclear safeguards applications based on the liquid organic scintillator EJ-309 with materials of different thickness surrounding the detector. The investigation was performed on the simulated data obtained by the MCNPX-PoliMi numerical code based on the Monte Carlo method. Among the various materials (polyethylene, iron, aluminum, and graphite) investigated as layers around the scintillator, polyethylene and iron have shown the most promising characteristics for evaluation of fast neutron energy spectra. The simulated pulse height distributions were summed up for each energy bin in the neutron energy range between 1 MeV and 15 MeV in order to obtain better counting statistics. The unfolded results for monoenergetic neutron sources obtained by a first order of Tikhonov regularization and non-linear neural network show very good agreement with the reference data while the evaluated spectra of neutron sources continuous in energy follow the trend of the reference spectra. The possible advantages of a novel spectrometer include a less number of input data for processing and a less sensitivity to the noise compared to the scintillation detector without surrounding materials.

This paper deals with the improvements of the linear artificial neural network unfolding approach aimed at accurately determining the incident neutron spectrum. The effects of the transfer functions and pre-processing of the simulated pulse height distributions from liquid scintillation detectors on the artificial neural networks performance have been studied. A better energy resolution and higher reliability of the linear artificial neural network technique have been achieved after implementation of the results of this study. The optimized structure of the network was used to unfold both monoenergetic and continuous neutron energy spectra, such as the spectra of 252Cf and 241Am-Be sources, traditionally used in the nuclear safeguards experiments. We have demonstrated that the artificial neural network energy resolution of 0.1 MeV is comparable with the one obtained by the reference maximum likelihood expectation-maximization method which was implemented by using the one step late algorithm. Although the maximum likelihood algorithm provides the unfolded results of higher accuracy, especially for continuous neutron sources, the artificial neural network approach with the improved performances is more suitable for fast and robust determination of the neutron spectra with sufficient accuracy.

Monte Carlo techniques were usually used to study the effect of scattered photons on a radiographic X-ray image, Such approach is accurate, but computer time consuming. On the other hand, the exposure buildup factors can be used as approximate and efficient assessment to account for the scattering of X-rays. This method uses the known radiography parameters to find the resulting detector exposure due to both scattered and un-collided photons. A model for radiography simulation, based on X-ray dose buildup factor, is proposed. This model includes non-uniform attenuation in voxelized object of imaging (patient body tissue). Composition of patient body is considered as a multi-layer structure. Various empirical formulas exist for multi-layer structure calculations and they all calculate multi-layer buildup factors by combining single-layer buildup factors. The proposed model is convenient in cases when more exact techniques (like Monte Carlo) are not economical. (C) 2009 The Japan Society of Applied Physics

The objective of the study presented in this paper is the analysis of influence of different buildup factor forms on a simulated radiographic image. Simulated radiographic images are obtained by means of the ray-tracing technique. Scattered photons are modelled using the generally accepted geometric progression form, linear form and tabulated data of buildup factors. Simulated images were compared to the reference results obtained by Monte Carlo calculation. The best agreement to Monte Carlo simulated images is achieved for the geometric progression form of buildup factor.

1861 Objectives Absorbed dose calculation must start with a prescribed dose but also must include cells dose calculations. Our integrated dosimetry system uses macro module for calculation of average tumor dose for specific patient and micro module for calculation of average cells doses of a specific tumor type. This system is used to calculate underdose/overdose of cells for tumor which received the prescribed dose. Methods Average tumor (sphere, R=5mm) dose and cell doses were calculated using Monte Carlo code FOTELP. The tumor model was randomly filled with cells (spheres, r=400µm, st.dev=50µm; volume density=40%). Simulations were done for different labelled cells densities (f=1,10,25,50,100%) and radionuclides I-131, Lu-177 and Y-90. Activities were homogenously distributed in cell volumes. Results Macro module was used to calculate the administration activity required to obtain the prescribed dose, which was applied in the micro module to calculate average absorbed doses of tumor cells presented as dose cells histograms. For I-131 and Lu-177 percentage of underdose cells was very high (90%) for f=1%, while for f=100% percentage of overdose cells was high (99%). In all cases less than 1.2% of cells received the prescribed dose. For Y-90 this percentage was 20.5%. Conclusions An integrated dosimetry system, which includes calculation of tumor absorbed dose for specific patient, and calculation of all tumor cell doses, must be used in preclinical radionuclide therapy studies as well as in clinical practice.

This paper proposes a neutron source identification method based on the spectral analysis of neutron pulse height distributions obtained with liquid scintillation detectors. The fact that shielded and unshielded neutron sources have clearly defined spectral components with specific locations and intensities offers the possibility of identifying the sources based on spectral features alone, without having to unfold the energy spectra. Analysis of simulated and experimental data confirms that this new identification method is promising, and that good resolution power can be achieved.

Three-dimensional point-kernel multiple scatter model for radiography simulation, based on dose X-ray buildup factors, is proposed and validated to Monte Carlo simulation. This model embraces nonuniform attenuation in object of imaging (patient body tissue). Photon multiple scattering is treated as in the point-kernel integration gamma ray shielding problems via scatter voxels. First order Compton scattering is described by means of Klein-Nishina formula. Photon multiple scattering is accounted by using dose buildup factors. The proposed model is convenient in those situations where more exact techniques, like Monte Carlo, are not time consuming acceptable.

A three-dimensional (3D) point-kernel multiple scatter model for point spread function (PSF) determination in parallel-beam single-photon emission computed tomography (SPECT), based on a dose gamma-ray buildup factor, is proposed. This model embraces nonuniform attenuation in a voxelized object of imaging (patient body) and multiple scattering that is treated as in the point-kernel integration gamma-ray shielding problems. First-order Compton scattering is done by means of the Klein-Nishina formula, but the multiple scattering is accounted for by making use of a dose buildup factor. An asset of the present model is the possibility of generating a complete two-dimensional (2D) PSF that can be used for 3D SPECT reconstruction by means of iterative algorithms. The proposed model is convenient in those situations where more exact techniques are not economical. For the proposed model's testing purpose calculations (for the point source in a nonuniform scattering object for parallel beam collimator geometry), the multiple-order scatter PSF generated by means of the proposed model matched well with those using Monte Carlo (MC) simulations. Discrepancies are observed only at the exponential tails mostly due to the high statistic uncertainty of MC simulations in this area, but not because of the inappropriateness of the model.

The objective of this study is the determination of the distribution of the activity of radio-elements contained in radioactive waste packages by means of single photon emission computed tomography (SPECT). A three- dimensional (3D) projector simulator for a parallel hole collimator and NaI(Tl) scintillator, based on the point kernel integration method, is proposed. The model takes into account the attenuation and scatter of gamma rays. Primary scattered photons are treated by the Compton process and Klein-Nishina formula, and the multiple scattering is accounted for by means of the dose buildup factor normally used in shielding problems. An advantage of the proposed model is that it offers the possibility of generating a full two-dimensional point-spread function (PSF) that can be used for 3D reconstruction. The developed model is convenient in those situations where more exact techniques (such as Monte Carlo simulation) are not economical. The model has been evaluated for the package having homogenous density with the Cs-137 point source inside.