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

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) arecalculated using radioactive decay equations within one year continues operation of reactor withthe neutron flux: 109n/cm2.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 (UAL4 with 90.2 %) within one year. In addition, the burnup of reactor’s fuel has been calculated based on the obtained results.

کلیدواژه ها:

Burnup ، Decay equations ، Mass ، Radioisotope ، Reactor