monte-carlo simulation

Scandium-43 (43Sc) is a positron emitter radioisotope that can be used in PET imaging by emitting 511 KeV (176%) annihilation photon and 372.76 KeV gamma ray (23%). In this study, the theoretical production yield of 43Sc and its accompanying impurities in the 43Ca(p,n), 44Ca(p,2n), 46Ti(p,a), 42Ca(d,n), 40Ca(a,p) and 41K(a,2n) reactions to produce this radioisotope in 30 MeV Karaj cyclotron were calculated by simulating the excitation functions of these reactions in TALYS-1/96 and EMPIRE-3-2-2 Monte Carlo codes, and the mass stopping power in SRIM-2013 Monte Carlo code and the numerical integration of the production yield formula in MATLAB in the energy range of 0 to 30 MeV, and were compared with the experimental results. 44Sc is a radioisotopic impurity that is produced in the 44Ca(p,2n) and 41K(a,2n) reactions and it’s practically impossible to remove it from the final product; it’s produced in the 46Ti(p,a) reaction and can be removed by determining the energy range; and it isn’t produced in the 43Ca(p,n), 42Ca(d,n) and 40Ca(α,p) reactions. The results show that for the production of pure 43Sc with the highest yield, 43Ca(p,n) is the best reaction for Karaj cyclotron; but to produce a pure product with a suitable yield and the lowest possible cost, the 40Ca(α,p) reaction is a better choice.

Protection against ionizing radiation has been a global concern for decades. In the field of radiation protection, photons and neutrons are particularly important, because these rays do not have an electric charge and have a high penetration depth in different materials. One of the most useful neutron sources is californium-252, which produces a very high flux of fast neutrons along with gamma photons. Considering the portability of californium-252, one of its advantages is the treatment of cancer with alpha particles by capturing its neutrons by boron-10. Therefore, it is necessary to design a shield with suitable materials and reduce the volume and weight of the shield in order to facilitate the moving of the source with the shield and to reduce the gamma and neutron dose. In this research, by using a pre-designed shield, the protective properties of kennertium, stainless steel, tungsten carbide and polyethylene/bismuth (78.5% wt.) to reduce gamma equivalent dose and magnesium borohydride, aluminum borohydride, zirconium borohydride, lithium borohydride, thorium borohydride, beryllium borohydride, premadex and sodium cyanoborohydride to reduce Neutron equivalent dose was simulated with the help of MCNPX Monte Carlo code. The simulation results indicated that magnesium borohydride with a thickness of 17.5 cm and kennertium with a thickness of 1 cm led to a 5.5% decrease in the total dose rate at a distance of 1 meter from the source and a 39% decrease in the shielding thickness compared to the reference article (the material and the thickness are stated in the text). Therefore, magnesium borohydride and kennertium were determined as the best neutron and gamma absorbers, respectively. In the continuation of this study, the development and investigation of other advanced materials are suggested as an optimal shielding in a mixed neutron and gamma field.

In this research, a study is performed in order to determine the optimal spectral filtration in dual-source imaging system and its effect on increasing the mean energy and energy separation between high and low energy spectra .X-ray spectrum is simulated using Monte carlo Fluka 2011 simulation code. X-ray tube with tungsten anode and 1.6 mm aluminum inherent filter was designed. Then, an additional filter was applied to the 140 keV tube and the low-energy spectrum was left unchanged to prevent further reductions in the output .In the next step,in order to minimizing the overlaps among the spectra, former filter was implemented and mean energy difrrences was investigated. Filter’s materials have selected from several elements in which, Tin and bronze are practical options. Based on our results, tin filter with a thickness of 0.5 mm has increased the average energy difference from 17.8 to 40.6 keV. In the second evaluation, using bronze filter on high and Eribium on the low energy tube, the average energy difference was increased from 17.8 to 45.87 keV.In our evaluation The overlap of low and high energy spectra have been decreased significantly.In addition to examine the sensitivity of different materials, the difrences in dual energy ratio has been compared. The larger the difference is, the contrast between two materials has increased .This increase in dual energy ratio is expected to improve the performance of dual-source CT scans in increasing contrast and improving image quality and dose reduction.

Radiation shielding materials and components of shielding structures should have good mechanical properties, long-term usability and appropriate flexibility. Nano composites have good micro structural and mechanical properties in different conditions. This article examines the principles of gamma ray shielding, including the interaction of gamma rays with matter, bismuth-based hetero junction nano composites by combining specific concentrations (90, 90, 85, 80 percent) of bismuth oxide with vanadium as a protective material for effective radiation protection. Gamma pays. Discusses various methods of evaluating gamma ray shielding, including measurement techniques and computer simulations using the Geant4 Monte Carlo tool. Also, to check the performance of these nano composites, we have calculated effective parameters in gamma ray attenuation, such as mass attenuation coefficient, mean free distance, one-tenth layer and accumulation coefficient. Finally, this article provides an overview of gamma ray shielding, which is used to ensure the safety of people and equipment against radiation.

The impacts of nanoparticles in radiation therapy have been investigated during the recent decades. The present study was conducted with the aim of investigating the effect of different physical interaction models on dose calculations using gold, hafnium and gadolinium nanoparticles. A nanoparticle with a diameter of 50 nm was simulated in a cubic water phantom and irradiated by protons with the energies of 5, 50 and 150 MeV using Geant4 Monte Carlo toolkit. The current study has considered various parameters, including the energy spectrum of secondary electrons and photons, radial dose distribution (RDD), dose enhancement factor (DEF), around the nanoparticle with three different materials and two physical interaction models. The obtained data showed that for gold nanoparticles, the Penelope model generated a greater number of secondary electrons than the Livermore model; however, for the other two nanoparticles, the Livermore model produced a greater number of secondary electrons than the Penelope model. In the RDD graphs, the Penelope model presents a 10% difference compared to the Livermore model up to a distance of 6 nm from the nanoparticle’s surface (along the radial axis in water). Furthermore, the Livermore model indicates a 16% and 10% increase in dose compared to the Penelope model up to a distance of 9 nm from the surface of hafnium and gadolinium nanoparticles, respectively. In the case of DEF, the dose deposited around the gold nanoparticle was increased by 14, the highest amount compared to DEF of hafnium and gadolinium nanoparticles which is 10 and 6, respectively

Bromine-76 (half-life = 16.2 hours) is a positron emitter radionuclide which has a high potential for use in nuclear medicine; but due to the difficulty of producing commercial quantities, it is only used in laboratory studies. This radionuclide is usually produced through the reaction of 76Se(p,n)76Br. The purpose of this research is to investigate the possibility of 76Br commercial production by bombarding targets made of stable germanium isotopes with 7Li+ heavy ion. The excitation functions of 70Ge(7Li+,n)76Br, 72Ge(7Li+,3n)76Br, 73Ge(7Li+,4n)76Br, 74Ge(7Li+,5n)76Br and 76Ge(7Li+,7n)76Br reactions were drawn using the EMPIRE and LISEcute++ codes, and from the comparison of these excitation functions, the 72Ge(7Li+,3n)76Br in the energy range of 30 to 40MeV was selected as the premier reaction. The maximum theoretical production yield in the energy of 40MeV for these codes are 32.46MBq/µAh and 61.43MBq/µAh, respectively. The theoretical and experimental yields of the 76Se(p,n)76Br at energy of 16MeV are 506.61MBq/µAh and 88MBq/µAh, respectively. From the comparison of the theoretical production yield of 72Ge(7Li+,3n)76Br and 76Se(p,n)76Br reactions, it can be concluded that the 72Ge(7Li+,3n)76Br reaction is considered only when the target of 72Ge or a combination of that have long-term irradiation capability (without melting) and thus produce more 76Br activity in practice.

‎In this paper, under adaptive hybrid progressive censoring samples, Bayes estimation of the multi-component reliability, with the non-identical-component strengths, in unit generalized Gompertz distribution is considered. This problem is solved in three cases. In the first case, strengths and stress variables are assumed to have unknown, uncommon parameters. In the second case,  it is assumed that strengths and stress variables have two common and one uncommon parameter, so all of these parameters are unknown. In the third case, it is assumed that strengths and stress variables have two known common parameters and one unknown uncommon parameter. In each of these cases, Bayes estimation of the multi-component reliability, with the non-identical-component strengths, is obtained with different methods. Finally, different estimations are compared using the Monte Carlo simulation, and the results are implemented on one real data set.