The Assessment of Radioisotopes and Radiomedicines in the MNSR Reactor of Isfahan and Obtaining the Burnup by Applying the Obtained Information

Whereas nowadays production and reprocessing the radionuclides are very significant and urgent, so in this paper according to operation of Isfahan Miniature Neutron Source Reactor (MNSR) that is a pool-tank reactor, the values of some radionuclides such as: ⁹⁹Mo and ^99mTc which are applied as radiomedicines and other radionuclides including fission products and actinides which are either fissile or fertile, within one year continues operation of reactor with the neutron flux: 10⁹n/cm².sec have been computed. In order to calculate the values of produced radionuclides the changes of nucleuses densities through all the differential equations of atoms density variation have been written and then by solving them by numerical method and also by using the MATLAB software, according to the kind of applied fuel and its enrichment percent (UAL₄ with 90.2 percent), the values of the mentioned produced radionuclides at the core of this reactor have been computed during one year. In addition based on the obtained results, the burnup of reactor´s fuel has also been defined.