Calculation of Produced Radioisotopes and Burn up in the Miniature Neutron Source Reactor Using Radioactive Decay Equations

ABSTRACT In this paper, the amounts of some produced radioisotopes such as radiomedicines and actinides, which are either fissile or fertile, in the Miniature Neutron Source Reactor (MNSR) are calculated using radioactive decay equations within one year continues operation of reactor with the neutron flux: 10 9 n/cm 2 .sec. In order to calculate the values of produced radioisotopes, the variations of nucleuses densities of radionuclides have been written through all the differential equations of atom densities variations, then the amounts of the produced radioisotpes at the core of this reactor have been computed by solving the mentioned equations through numerical method and also using the MATLAB software, according to the type of applied fuel and its enrichment percentage (UAL 4 with 90.2 %) within one year. In addition, the burnup of reactor's fuel has been calculated based on the obtained results

Design and Simulation of a New Model for Treatment by NCT

In this investigation, neutron capture therapy (NCT) through high energy neutrons using Monte Carlo method has been studied. In this study a new method of NCT for a sample liver phantom has been defined, and interaction of 12 MeV neutrons with a multilayer spherical phantom is considered. In order to reach the desirable energy range of neutrons in accord with required energy in absence of eligible clinical neutron source for NCT, this model of phantom might be utilized. The neutron flux and the deposited dose in the all components and different layers of the mentioned phantom are computed by Monte Carlo simulation. The results of Monte Carlo method are compared with analytical method results so that by using a computer program in Turbo-Pascal programming, the deposited dose in the liver phantom has been computed

Presenting and simulating an innovative model of liver phantom and applying two methods for dosimetry of it in neutron radiation therapy

AimA new model of liver phantom is defined, then this model is simulated by MCNPX code for dosimetry in neutron radiation therapy. Additionally, an analytical method is applied based on neutrons collisions and mathematical equations to estimate absorbed doses. Finally, the results obtained from two methods are compared to each other to justify the approach.BackgroundThe course of treatment by neutron radiation can be implemented to treat cancerous tissues, although this method has not yet been widespread.The MIRD and the Stylized Family Phantom were the first anthropomorphic phantoms, although the representation of internal organs was quite crude in them. At present, a water phantom is usually used for clinical dosimetry.Materials and methodsEach of the materials in an adult liver tissue including water and some organic compounds is decomposed into its constituent elements based on mass percentage and density of every element. Then, the accurate mass of every decomposed material of human liver tissue is correlated to masses of the phantom components.ResultsThe absorbed doses are computed by MCNPX simulation and analytical method in all components and different layers of this phantom.ConclusionsWithin neutron energy range of 0.001[[ce:hsp sp="0.25"/]]eV–15[[ce:hsp sp="0.25"/]]MeV, the calculated doses by MCNPX code are approximately similar to results obtained by analytical method, and the derived graphs of both methods approve one another. It is also concluded that through increasing the incident neutron energy, water receives the largest amounts of absorbed doses, and carbon, nitrogen and sulfur receive correspondingly less amounts, respectively