mcnpx code

The feasibility of producing the iridium-192 radioisotope has been investigated in this work. Iridium-192 has various applications in nuclear medicine, particularly in brachytherapy. To increase the production yield and the high purity of this product, the direct production parameters of this radioisotope were studied based on the 191Ir (n,γ) 192Ir reaction. It is crucial to determine the appropriate irradiation location according to the neutron flux and the cross-section of the neutron reaction with the target material. By analyzing the experimental spectrum of neutrons exiting the reactor, as well as simulating the reactor using the MCNPX code, detailed information was obtained on the flux and spectrum of thermal neutrons and the irradiation sites. Calculations related to the flux of thermal neutrons were performed both theoretically and through simulation. Considering the information obtained from the simulation and comparing it with the experimental diagram of the reaction cross-section, the optimal irradiation site for the production of iridium-192 was determined. By comparing the simulation results related to the final activity at different reactor locations, the best site for producing iridium-192 with the highest activity was also identified.

Gamma ray measurement in various research fields requires high efficient detectors. In photon dosimetry, NaI(Tl) scintillation detector as one of the inorganic scintillation detector is noticeable, due to have the high amount of light output. In this study, the basics determination of photon dosimetry for the NaI(Tl) scintillation detector utilizing the Monte Carlo code (MCNPX) and using different methods of dose calculation (tally F6, * F4, + F6 and * F8) is studied. Regularly, the output of a radiation detector (counting the number of pulses) cannot be used to determine the radiation dose value. Therefore, in this study the spectro-dosimetry method based on software method is used to find out the value of the conversion coefficients to convert the detector spectrum to the value of air karma. In this method, the radiation dosimetry response is obtained with use of the MCNPX code simulation. The response function of the NaI(Tl) 3"×3" scintillation detector for several specific gamma rays was determined and then the functions of energy dependent conversion coefficients for calculating the dose values were obtained. Finally, with comparison of the measured data and simulation calculations results it is shown that the proposed method has a high accuracy in photon dosimetry

Production of semiconductors such as silicon doped with phosphorus has many applications in the production of electronic components and various industries such as aerospace. The process of impurity making, which is called silicon doping, can be done by both chemical and nuclear methods. Since the uniformity of the injected impurities is not suitable in the chemical method, the methods of silicon doping by neutron irradiation method are strongly followed in the world. In this work, the potential of the thermal column of Tehran Research Reactor for silicon doping is investigated using MCNPX simulation code. The results show that the thermal neutron flux as well as the ratio of thermal to fast neutron flux in the optimal location are 1.2×1012 n/s.cm2 and 441, respectively, which shows that the thermal column of the Tehran research reactor can be a suitable place for silicon doping.

In the present work, dose equivalents of radiosensitive organs in head and neck region have been calculated during nasopharynx proton therapy. For this purpose, a middle-size tumor was designed with the use of clinical information, and was then incorporated into the adult male ICRP phantom. According to the medical data, eight radiation paths around the neck were selected. After that, appropriate proton beams for each radiation paths were designed considering the type and thickness of tissues in the beam path. Since both of soft tissue and bone are existed in pathways of radiation, firstly, the effect of the existence of bone tissue in the path of protons on the position of the Bragg peak was studied. Results showed that when the thickness of bone tissue within the phantom increased by 1 cm, the Bragg peak was pulled back about 0.6 to 0.8 cm. It was also found that the displacement of the Bragg peak by increasing the bone thickness follows a polynomial function for each proton energy. Considering the thickness of the tumor, optimized SOBP were designed for each of eight directions. Finally, doses to different sensitive organs of head and neck region were computed in terms of the therapeutic dose of the tumor. Results indicated that thyroid and brain received higher doses in comparison with other organs.

One of the most important groups that can be exposed to nuclear radiation is pregnant women and their fetus. This group of people can be technicians at nuclear centers located in the patients undergoing treatment or nuclear medicine diagnostic centers. The purpose of the paper is to determine the rate of doses received by female technicians who are indirectly in the vicinity of nuclear radiation. In order to verify the results of computer simulation, a comparison was made with experimental results from the nuclear medicine department of Imam Hossein Hospital in Shahroud. The statistical analysis of the results showed that there was a significant difference between the simulation results and there are no experimental results.

Brachytherapy, is a method to treat prostate cancer in which a radiation source is placed inside or next to the cancer affected tissues. Purpose of this article is to determine the uncertainty level of received dose in prostate tissue due to the relocation of the placed radiation sources and inflation in prostate after positioning the radiation seeds. To simulate the model 6711 Amersham (activity level equal to 0.5 mCi) iodine-125 brachytherapy source, we used MCNPX 2.6 code and TG-43U1 protocol, and we used ORNL phantom. Treatment plan included 76 sources consisting iodine-125 radioactive, once placed in the form of seeds and once applied in the form of beads in three initial volume: 30.02, 38.01 and 52.01 cm3 and three different time steps were placed in prostate phantom. Three-time steps were investigated, including the moment after placing, the zero days and thirty days after placing. In the first step, the obtained dose of radiation in three different prostate volumes, for both seed distribution and bead distribution in healthy and cancerous tissue, was calculated. In the second step, position of the sources was relocated from the assumption, and causes a change in dose distribution. Considering the movements in 3 directions of left-to-right (1.8 mm), intra-exterior (2.1 mm), up-down (3.4 mm) after implantation the equivalent dose was calculated. The maximum uncertainty of received dose in a prostate with seed implanted sources and the volume of 52.01 cm3 and it was reported as +20%. At the third step, in addition to the change in the location of sources, inflation rate of 12% was observed in comparison with the primary volume at the first step of treatment process which resulted in a smaller dose rather than the previous stage.In this treatment plan, relocating of the sources and 12% inflation of the tumor were analyzed. Uncertainty of source relocating was calculated as ±20% in zero-day of positioning. Also, 12% inflation of the tumor and source relocation after 30 days, in a prostate with the volume of 58.25 cm3 curing by point sources, shown 21.04% reduction in received dose for healthy and cancer tissues since the moment of placing. The maximum error of MCNPX code was calculated 0.03%.

Radiation detection is essential for determining of radiation dose. Depend on the detector and dosimetry method, detection process is performed in different levels. Pulse counting is the first level of detection. Typically, the output of a radiation detector for determining value of the radiation dose cannot be used directly. Through changing the response function or the readout detector, is trying to create relevance and proportionality between detector readout and value of the dose equivalent. For this purpose, there are various software methods and hardware methods are obtained dose equivalent values. Software methods such as convolution, Deconvolution, etc. and hardware methods such as additional of layers of modulators, etc have been used. Selective data sampling of multi-channel energy is a software method which has been studied in this paper. The main purpose of this method, is obtaining the response of photon dosimetry at different energies. The response function of NaI(TL) 3”×3” detector for 0.411 MeV to 3 MeV gamma-rays has been simulated by using MCNPX code, and for calculating dose, the calibration coefficients were determined. Finally, the response energy of the detector has been drawn for photon dosimetry of the channels with multiple energy range. Results show that presented dosimety method has high level precision.

In this study, distribution of dose rate around the nuclear gauge device MC-1DR which located in shahrekord university was simulated by MCNPX code and was compared whit the measured values. Due to the asymmetry of device and neutron and gamma source positions, the dose rates were determined at a distance of 5, 30 and 100 cm in different directions. Base on the complex geometry of the inside of device, there are discripency between measured and simulated results in the some points. In general, the values show when the gamma source is positioned in safe mode. The maximum and minimum of dose rate are in below and back of the device. Also, in the left side neutron dose and in the right side gamma dose is greatest. Finally, for safe operating one hour is at most recommended at a distance of 1m in compare with standard threshhold, 12mrem per day.

Recently, a new Neutron Radiography (NR) beam line has been designed, constricted, installed and tested based on the use of E-beam tube of Tehran Research Reactor (TRR). Initial tests have been shown that the system can be used for different samples and purposes such as nuclear plates and rods fuels.  For this end the system need a suitable irradiation room which should be installed at the NR beam port. The present work is the analyses of the proposed irradiation room in view of irradiation safety. To do this, using MCNPX Monte Carlo code the proposed room structure and neutron beam catcher have been simulated and then neutron/gamma dose have been calculated at different locations around the room walls. The results show that the designed room and beam catcher have good performance in case of reduce the total neutron and gamma dose to 10 μSv/h which is desired value in view of radiation protection issue.