Decommissioning plan of research reactor first stage is the transport of the spent fuel from the reactor site to dry storage or to country of origin. Consequently, the radiological characterization for spent nuclear fuel is necessary for shielding calculation and dose rate estimation on the transport cask surface. It is the basis for safeguards, radiation protection and safety of personnel, public and the environment. It support the identification of contamination and assessment of potential risks. Most of the comprehensive characterization activities are conducted during the transition stage in preparation for decontamination and dismantling activities. Several methods were utilized for spent fuel inventory calculation and activation products in structure materials of a nuclear reactor.
In this paper concentrations and time- dependent activities of the source term of OPAL research reactor spent fuel will be calculated using MCNPX 2.7 Monte Carlo code. Calculations based on operational irradiation history in cycle 7, initial enrichment, discharge BU and neutron cross-section data library. Nuclides composition can be changed during cooling times as different decay schemes lead to production and destruction of nuclides. Also, radiation protection and safety management for the workers, public and for the environment in case of normal and emergency exposure during decommissioning activities will be evaluated based on the international basic safety standards.