mcnp4c code

The dosimetry of the brachytherapy sources is performed in the water source medium according to protocol TG-43U1. To achieve a resonable results on treatment, the use of the water material for all body tissues can be one of the faulty sources in delivering the correct dose to the tumor. Thus, we have focused on the dosimetry parameters in fat density 0.95 gr/cm3 and the muscle with density 1.05 gr/cm3 in the present study. The results are compared with the results of the water phantom to show the accurate accessment via the differences of the results.
Dissymmetry simulations for determining the parameters of the radial dose function and the anisotropy function in fat tissues, muscle and water phantom distances and different angles has been using MCNP4C code.
The greatest relative differences of the phantom radial dose function of the fat tissue at distances below 1cm approximately reached 13%; and this rate increased, when the distance from the source increased; whereas at the distance of 5cm from the source, it approximately reached 167%. In the muscle tissue, the relative difference of these parameters was about 3% at the distance of 0.5cm, while at the distance of 5cm from the source, it approximately increased 16%. The maximum relative difference of the anisotropy parameter of the fat and muscle tissue phantom compared with the water were observed more than 2% and 3%, respectively.
In the clinical application of the 103Pd brachytherapy source, which is contracted in the treatment of the adjacent malignant tumors to the fat and muscle tissues, the correct decisions must be applied on the tissue dosimetry parameters in the treatment planings according to the tables of 3, 4, 5 and 6 in this study.

Performing successful BNCT experiments needs a suitable neutron source. Important factors of the neutron beam are flux and energy that are very important in the selection of neutron source. In most centers that use this method for treatment, reactor is a neutron source, which according to characteristics of the reactor appropriated neutrons are very high. High cost of constructing a BNCT center with using of reactor caused seeking other sources such as accelerator indirectly and radioisotope source directly that each has their own advantage and disadvantages. In this paper we created neutron beam by analysis Am-Be neutron source, using neutron filter technique and suitable moderators. The advantages of Am-Be neutron source are being inexpensive, easy portability, small size and well-designed shields. Therefore, by analyzing radioisotope neutron sourcesand Am-Be neutron source specially, we can prepare possible analysis radioisotope neutron source at boron neutron capture therapy. We hope to achieve suitable results by more studies. Neutron beam in 1keV energy created with using Am-Be neutron source and designed suitable neutron filter with using neutron absorbent materials that it will be used in testing BNCT. By studying and Identifying various materials such as oxides Alumina, graphite and beryllium as a moderator and materials such as boron, cadmium and titanium as absorbent materials to a cylindrical crust in filter has been used. Neutron Filter has been designed in the investigation of two parts. The first is consisting of a moderator with high scattering and very low percent and it is caused the fast Neutron servant brought back his spectrum Am-Be source in this without mono-energetic to the low energy transferred spectrum. Part II filter is consisting of the elements of boron, cadmium and titanium that are absorbent neutron with various energy, therefore they can exchange these neutrons in certain energy to mono-energetic. More analysis, study and designing suitable neutron conductors for increase neutron flux is recommended. Neutron filter passes neutron with energy 1keV that can be used in the BNCT experiments. According to data obtained from the implementation MCNP4C code, a peak is obtained in energy 1keV that indicate area under the flux 2.22E-05 n/cm2.s with error 0.0065 for a neutron. Flux obtained can be multiplied at the Am-Be source of power that is equal 108n/cm2.s until the total flux to be achieved. The total flux is obtained 2.22E+03n/cm2.s at 1 cm2. We must multiply total intensity at total area to achieve total neutron flux, Since the flux required for the BNCT experiments is 5*108n/cm2.s with using different ways and designing suitable reflectors and conductors, this neutron flux will be provided. This paper analyzed possible use of radioisotope neutron source by simulation Am-Be neutron source. We can solve many problems that exist for reactor source paying attention to characteristics of radioisotope sources such as being inexpensive, easy portability and small size also more studies are present in this base. Of course, with completing this simulation, we can be hopeful for practicality and remedy of patients in Iran.

90Sr/90Y source has been used for the intravascular brachytherapy to prevent coronary restenosis in the patients who have undergone angioplasty. The aim of this research is to determine the dose distribution of 90Sr/90Y source in a water phantom. In the present work, MCNP code has been applied to calculate the dose distribution around a 3 cm length of 90Sr/90Y source in a 30×30×30 cm3 water phantom. Also, the exact geometry of the source has been used in this simulation. Tally *F8:e which is suitable for beta ray dosimetry has been evaluated with less than %5 relative error in a sphere having 0.2 mm radius. The isodose curve for 10, 20, 40, and 90% depth dose (PDD) were derived based on the calculated dose curves along the parallel and perpendicular axis to the source.Discussion and The results obtained in this work are in a good agreement with the experimental result published by Buckley et al. and the International Atomic Energy Agency (IAEA) report in a water phantom. Therefore, the result of this research can be used in the intravascular brachytherapy.

In the present research, by the use of Monte Carlo method the TG-43 dosimetry parameters of 125I model 6711 have been determined and its relative dose has also been calculated. 125I model 6711 brachytherapy source which is manufactured by Amersham Company and has been approved by the American Association of physicists in Medicine (AAPM) was used in this simulation. The Source was considered in the center of 30 cm×30 cm × 30 cm water and soft tissue phantom and the relative dose variation along the parallel and perpendicular direction to the source axis were calculated using MCNP 4C code. Water having a density of 1 gram per cubic centimeter composed of two hydrogen atoms and one oxygen atom and soft tissue with density of 1.04 gram per cubic centimeter with accurate composition was used in this simulation. In this research, the percentage depth dose (PDD) variation along the different axis parallel and perpendicular to the source has been calculated using F6: p tally with less than 5% relative error. These data were then used to derive the isodose curves for PDD= 125%, 100%, 75%, 50% and 25%. Also, anisotropy function F(r,θ) and radial dose function g(r) have been calculated for the source.Discussion and The percentage dose variation and dosimetry parameters have been calculated in this research are in good agreement with the results of others. This finding can be used to improve brachytherapy result with this source. Since the absorbed dose near a source is high and its gradient is significant, only Monte Carlo simulation can be used for an accurate calculation. The results demonstrate a useful approach using MCNP code in dose calculation that can be applied in other fields of medical physics.